Text

Response to Request for Additional Information Regarding Review of the License Renewal Application - Set 1
	ML23193A846
Person / Time
Site:	Comanche Peak
Issue date:	07/12/2023
From:	John Lloyd; Luminant, Vistra Operating Co. (VistraOpCo)
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23193A845	List: CP-202300285, Response to Request for Additional Information Regarding Review of the License Renewal Application - Set 1;
References
	CP-202300285, TXX-23044 WCAP-10528-NP, Rev 3
	Download: ML23193A846 (1)
	v • d • e

Enclosure 3 contains Proprietary Information Withhold from Public Disclosure in accordance with 10 CFR 2.390 Comanche Peak Jay Lloyd Nuclear Power Plant Senior Director, Engineering (Vistra Operations

& Regulatory Affairs Company LLC)

P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 T 254.897.5337 CP-202300285 TXX-23044 July 12, 2023 U. S. Nuclear Regulatory Commission Ref 10 CFR 54 ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NUMBERS 50-445 AND 50-446 FACILITY OPERATING LICENSE NUMBERS NPF-87 and NPF-89 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING REVIEW OF THE LICENSE RENEWAL APPLICATION - SET 1

REFERENCES:

1. Letter TXX-22077, from Steven K. Sewell to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application, October 3, 2022 (ADAMS Accession No. ML22276A082)

2. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

3. Letter TXX-23022, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 2, April 24, 2023 (ADAMS Accession No. ML23114A377)

4. Electronic Communications, from A. Siwy (NRC) to K. Peters (Vistra), "Comanche Peak LRA -

Request for Additional Information - Set 1," June 14, 2023 (ADAMS Accession Nos.

ML23167A022 and ML23167A023)

Dear Sir or Madam:

In Reference 1, as supplemented by References 2 and 3, Vistra Operations Company LLC (Vistra OpCo) submitted a license renewal application (LRA) for the Facility Operating Licenses for Comanche Peak Nuclear Power Plant (CPNPP) Units 1 and 2. The NRC issued Requests for Additional Information (RAIs) to Vistra OpCo via Reference 4. Vistra OpCos responses to these RAIs are provided in Enclosure 1 of this letter. Enclosure 2 provides the non-proprietary Westinghouse Report WCAP-10528, Revision 3, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Comanche Peak Units 1 and 2 for the License Renewal Program (60 Years)." Enclosure 3 provides the proprietary Westinghouse Report WCAP-10527, Revision 3, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Comanche Peak Units 1 and 2 for the License Renewal Program (60 Years)." Enclosure 4 provides the Westinghouse Electric Company LLC (Westinghouse) Application for Withholding Proprietary Information from Public Disclosure and accompanying Affidavit. As Enclosure 3 contains information proprietary to Westinghouse, it is respectfully requested that Enclosure 3 be withheld from public disclosure in accordance with 10 CFR 2.390.

TXX-23044 Page 3 of 3 Enclosure 1 Index Enclosure CPNPP CPNPP LRA Request for Additional Information Topics No. LRA RAI 1.A B.2.2.1-1 High-Energy Line Break Analyses ASME Section III, Class 1 Fatigue Analysis of Piping, Piping Components, 1.B 4.3.2-1 and Equipment Environmentally Assisted Fatigue (Stress Analysis Method Rankings and 1.C 4.3.4-1 Sentinel Location Removal Method)

Environmentally Assisted Fatigue (Sentinel Location Identification and 1.D 4.3.4-2 Fatigue at Welds)

Environmentally Assisted Fatigue (Accumulator Nozzle Crotch Region 1.E 4.3.4-3 (180° Loc) 60-Yr Projected CUFen Discrepancy) 1.F 4.3.4-4 Environmentally Assisted Fatigue (TLAA Disposition Discrepancy) 1.G 4.3.5-1 Reactor Vessel Internals Fatigue Analyses 1.H 3.3.2.8c-1 Aging Management of Aluminum Components Exposed to Outdoor Air 1.I 4.7.1-1 Leak-Before-Break 1.J B.2.3.3-1 Reactor Head Closure Stud Thread Damage 1.K B.2.3.3-2 Reactor Head Closure Stud Thread Damage Corrective Actions 1.L B.2.3.3-3 Reactor Head Closure Stud Tension 1.M B.2.3.3-4 Reactor Head Closure Stud AMP Enhancements

Enclosure 1:

Responses to RAIs 37 pages follow

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.2.1-1 TXX-23044 and CP-202300285 Enclosure 1.A Page 1 of 5 LRA Section: 4.3.6, High-Energy Line Break Analyses NRC RAI No: B.2.2.1-1 (Postulation of non-Class 1 HELB locations for AMP)

Regulatory Basis:

Pursuant to 10 CFR 54.21(a)(3), the license renewal application (LRA) must demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation.

Background:

In its supplement dated April 6, 2023, the applicant revised the enhancement regarding the corrective actions program element of the Fatigue Monitoring aging management program (AMP). The revised enhancement indicates that the corrective action should consider the impact on high-energy line break (HELB) locations. In relation to this enhancement, LRA Section 4.3.6 addresses the time-limited aging analysis (TLAA) on the HELB analysis. LRA Section 4.3.6 explains that the HELB TLAA uses the screening criterion of a cumulative usage factor (CUF) value of 0.1 for break location postulation.

In comparison, Final Safety Analysis Report (FSAR), Section 3.6B.2 describes the current licensing basis (CLB) screening criteria that are used to determine the intermediate locations of postulated breaks for the HELB analyses. Specifically, FSAR Section 3.6B2.1.2 indicates that the CUF value of 0.1 is included in the screening criteria for HELB location postulation for ASME Code Section III, Class 1 piping.

FSAR Section 3.6.B2.1.2 also indicates that the postulation of HELB locations for non-Class 1 piping is, in part, based on the allowable stress range for expansion stress (SA),

consistent with Branch Technical Position MEB 3-1 (ADAMS Accession No. ML052340555).

SA may need to be adjusted by a stress range reduction factor that is determined by the number of thermal cycles, as addressed in the implicit fatigue analysis in LRA Section 4.3.3 that was dispositioned in accordance with 10 CFR 54.21(c)(1)(iii).

Issue:

The postulation of non-Class 1 HELB locations is not clearly discussed in LRA Section 4.3.6 and the revised enhancement of the Fatigue Monitoring AMP. Therefore, the staff needs clarification on whether the corrective actions in the revised enhancement will consider the potential impact of transient cycles on both Class 1 and non-Class 1 HELB locations (e.g.,

potential need for identification of additional HELB locations and related evaluation).

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.2.1-1 TXX-23044 and CP-202300285 Enclosure 1.A Page 2 of 5 Request:

Clarify whether the corrective actions in the revised enhancement will consider the potential impact of transient cycles on both Class 1 and non-Class 1 HELB locations. If not, explain why the corrective actions do not need to consider such impact on both Class 1 and non-Class 1 HELB locations.

Luminant Response:

The corrective actions in the revised enhancement for the fatigue monitoring AMP in Reference 1, Attachment W1 applies to all HELB (both Class 1 and non-Class 1) locations.

References:

1. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.2.1-1 TXX-23044 and CP-202300285 Enclosure 1.A Page 3 of 5 Associated LRA Revisions:

LRA Appendix A, Table A-3 (page A-50), Table A-3 is revised as follows:

Table A-3 List of LR Commitments and implementation Schedule No. Aging NUREG-1801 Commitment Implementation Management Section Program or Activity 1 Fatigue X.M1 Continue the existing Fatigue Monitoring AMP, including No later than 6 months prior Monitoring enhancements to: to the PEO, i.e.:

(A.2.1.1) a) The program will be modified to include EAF analyses for U1: 08/08/2029 locations, in addition to those listed in NUREG/CR-6260, U2: 08/02/2032 that are determined to be sentinel locations through the EAF screening evaluation. or no later than the last b) The program will be modified to monitor the CUFen at the refueling outage prior to the sentinel locations consistent with the supporting PEO.

environmentally assisted fatigue analyses.

c) The program will be modified to monitor the dissolved oxygen through the primary water chemistry program to ensure it will remain consistent with that assumed in the environmentally assisted fatigue analyses.

d) The program will be revised to account for additional critical thermal and pressure transients for components that have been identified to have a fatigue TLAA, as appropriate.

Critical transients are those that require monitoring to ensure the CUF/CUFen remain < 1.0. Examples of why a transient would not be monitored is if it results in stresses below the endurance limit or occurs with an already counted transient.

e) The program will be modified to include acceptance criteria based on the 60-year cycle projections used in the

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.2.1-1 TXX-23044 and CP-202300285 Enclosure 1.A Page 4 of 5 supporting environmentally assisted fatigue analyses to ensure that the CUFen does not exceed 1.0.

f) The program will be modified to provide clarity on when to initiate corrective action. These corrective actions may include repair of the component, replacement of the component, a more rigorous fatigue analysis, or a flaw tolerance analysis consistent with ASME XI, Appendix L.

The corrective action should consider the impact on all (both Class 1 and non-Class 1) HELB locations and ASME Section III, Class 2 & 3 allowable stress analyses.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.2.1-1 TXX-23044 and CP-202300285 Enclosure 1.A Page 5 of 5 LRA Appendix B, Section B.2.2.1, Fatigue Monitoring (page B-21) is revised as follows:

Element Effected Commitment

7. Corrective The program will be modified to provide clarity on when to Actions initiate corrective action. These corrective actions may include repair of the component, replacement of the component, a more rigorous fatigue analysis, or a flaw tolerance analysis consistent with ASME XI, Appendix L. The corrective action should consider the impact on all (both Class 1 and non-Class 1) HELB locations and ASME Section III, Class 2 & 3 allowable stress analyses.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.2-1 TXX-23044 and CP-202300285 Enclosure 1.B Page 1 of 2 LRA Section: 4.3.2, ASME Section III, Class 1 Fatigue Analysis of Piping, Piping Components, and Equipment NRC RAI No: 4.3.2-1 (RCPs and SGs Components and Locations Subject to Fatigue Waiver Evaluations)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

In its supplement dated April 6, 2023, the applicant revised LRA Section 4.3.2 with respect to the components that are subject to fatigue waiver evaluations. Specifically, the applicant indicated that reactor coolant pump (RCP) and steam generator (SG) locations conform to the waiver of fatigue requirements of ASME Code,Section III, Subparagraph NB-3222.4(d).

Issue:

The applicants supplement does not describe the specific locations or components of the reactor coolant pumps (RCPs) and steam generators (SGs) that are subject to the fatigue waiver evaluations.

In addition, LRA Table 4.3.4-1 identifies the RCP casing to discharge nozzle interface as a limiting environmentally assisted fatigue (EAF) location. Therefore, the staff needs clarification on whether the fatigue wavier is applied to the RCP casing but not to the casing to nozzle interface.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.2-1 TXX-23044 and CP-202300285 Enclosure 1.B Page 2 of 2 Request:

1. Describe the specific locations or components of the RCPs and SGs that are subject to the fatigue waiver evaluations.

2. Considering the identification of the RCP casing to discharge nozzle interface as a limiting EAF location in LRA Table 4.3.4-1, clarify whether the fatigue wavier is applied to the RCP casing but not to the RCP casing to nozzle interface. As part of the response, describe the specific component of the RCP casing to discharge nozzle interface location (e.g., RCP casing to discharge nozzle weld).

Luminant Response:

1. Waivers of fatigue requirements per ASME Code,Section III, Subsection NB-3222.4(d) are credited in the current licensing basis fatigue evaluations for the following RCP and SG locations: Unit 1 RCP Thermal Barrier Flange, Unit 1 RCP Casing Large Feet, Unit 1 and Unit 2 RCP No. 1-3 Seal Leakoff nozzles, Unit 1 and Unit 2 RCP No. 3 Seal Injection nozzle, Unit 2 RCP Casing, Unit 2 RCP Main Flange, Unit 2 RCP Weir Plate, Unit 2 Casing Suction Nozzle, Unit 1 and Unit 2 SG tube weld plug and tube plug.

2. In the current licensing basis for the RCP, some locations credit ASME Code,Section III, Subsection NB-3222.4(d) waiver of fatigue requirements (see item 1) and others include explicit CUF values. For Unit 1, neither the RCP casing nor the RCP casing to nozzle interface apply a fatigue waiver. For Unit 2, the RCP casing credits a fatigue waiver but the RCP casing to nozzle interface does not. The RCP casing to nozzle interface, which is included as a sentinel location in the LRA Table 4.3.4-1, includes explicit CUF values for Unit 1 and Unit 2. The specific location is described in the current licensing basis evaluation as the location where the discharge nozzle transitions into the pump casing. There is no discharge nozzle to pump casing weld at this location. In summary, all fatigue locations included in the current licensing basis for the RCP are addressed either by fatigue waivers or explicit CUF values demonstrated in LRA Section 4.3.2 to be applicable through the PEO as amended by Reference 1.

References:

1. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-1 TXX-23044 and CP-202300285 Enclosure 1.C Page 1 of 3 LRA Section: 4.3.4, Environmentally Assisted Fatigue NRC RAI No: 4.3.4-1 (Stress Analysis Method Rankings and Sentinel Location Removal Method)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

LRA Section 4.3.4 addresses the TLAA on environmentally assisted fatigue (EAF) for ASME Code Section III, Class 1 pressure boundary components and piping. LRA Section 4.3.4 indicates that, in the detailed evaluation of EAF for sentinel (limiting) locations, the applicant considered the technical rigor of different stress analysis methods and the level of conservatism associated with the stress analysis methods. The LRA also explains that the results of determining the technical rigor and the associated conservatism are the stress analysis method rankings for EAF locations (also called stress basis comparison rankings).

The LRA indicates that EAF locations with the lower screening environmental cumulative usage factor (CUFen) values and lower rankings may be removed from the sentinel location list in comparison with the other EAF locations.

Issue:

However, the LRA does not clearly discuss the stress analysis method rankings and their technical bases. In addition, the staff found a need to clarify whether EAF locations are removed from the sentinel location list only if both the screening CUFen value and stress analysis method ranking are lower than those of a more limiting location, respectively.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-1 TXX-23044 and CP-202300285 Enclosure 1.C Page 2 of 3 Request:

1. Describe the stress analysis method rankings and their technical bases.

2. Clarify whether EAF locations are removed from the sentinel location list only if both the screening CUFen value and stress analysis method ranking are lower than those of a more limiting location. If not, provide justification for why the EAF locations with the higher screening CUFen or higher stress analysis method ranking may be removed from the sentinel location list.

Luminant Response:

1. The stress analysis method rankings (called stress basis comparison rankings in Section 4.3.4 of the LRA) are defined to represent the relative degree of conservatism included in the analysis to provide a consistent basis of comparison, ordered numerically from most conservative to least conservative (e.g., 1 to 3). For example, a comparison ranking of one would involve simplified or one-dimensional analysis, a ranking of two would consider an elastic finite element analysis, and a ranking of three would consider a plastic finite element analysis. The rankings for 1-3 are generally for equipment while piping has similar rankings but from 1-5. The following illustrates how the rankings are applied in the comparisons:

- If Component A has a higher CUFen and has a higher stress basis comparison ranking than Component B, then Component B can be eliminated from consideration.

- If Component A has a higher CUFen than Component B, and Component A considered the same stress basis comparison ranking as Component B, then Component B can be eliminated from consideration.

- If Component A has a higher CUFen than Component B but has a lower stress basis analysis ranking than Component B, then both Component A and Component B must be considered as potential sentinel locations. Further, more detailed evaluations would be necessary to identify the true sentinel location between Component A and Component B.

The technical basis for the stress basis comparison methodology is generally explained in Section 4.1.2 of EPRI Technical Report 3002018262, Revision 1, Environmentally Assisted Fatigue Screening Methods, Reference 1.

2. Within a transient section, EAF locations are removed from the sentinel location list only if both the screening CUFen value is lower than that of a more limiting location and the stress analysis method ranking is the same or lower than the more limiting location.

That is, between two screening CUFen values with the same level of technical rigor, the

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-1 TXX-23044 and CP-202300285 Enclosure 1.C Page 3 of 3 lower CUFen value would be screened out within a given transient section and material category.

References:

1. EPRI Technical Report 3002018262, Revision 1, Environmentally Assisted Fatigue Screening Methods.

Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-2 TXX-23044 and CP-202300285 Enclosure 1.D Page 1 of 2 LRA Section: 4.3.4, Environmentally Assisted Fatigue NRC RAI No: 4.3.4-2 (Sentinel Location Identification and Fatigue at Welds)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

LRA Table 4.3.4-2 describes the sentinel locations of EAF for piping lines and the associated 60-year projected CUFen values. The table identifies only stainless steel as a fabrication material for the limiting piping locations of EAF and does not include carbon steel, low alloy steel, or nickel alloy.

Issue:

In contrast, LRA Section 4.3.4 indicates that the sentinel location is identified for each material type in a given transient section. The transient section is a group of sub-components and locations that experience the same transients. In addition, LRA Section 4.7.1 indicates that the reactor vessel nozzle welds are fabricated with nickel alloys.

However, it is unclear to the staff whether the limiting EAF locations in LRA Table 4.3.4-1 (equipment EAF) or 4.3.4-2 (piping EAF) are bounding for the nickel alloy welds of the reactor vessel nozzles in terms of CUFen and environmental fatigue correction factor (Fen) in the EAF analysis.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-2 TXX-23044 and CP-202300285 Enclosure 1.D Page 2 of 2 Request:

1. Justify why carbon steel, low alloy steel, or nickel alloy locations are not identified as a limiting EAF location in LRA Table 4.3.4-2.

2. Clarify whether the limiting EAF locations in LRA Table 4.3.4-1 or 4.3.4-2 are bounding for the nickel alloy welds of the reactor vessel nozzles in terms of CUFen and environmental fatigue correction factor (Fen) in the EAF analysis. If so, describe the bounding locations and their CUFen and Fen values in comparison with the CUFen and Fen values of the nickel alloy welds of the reactor vessel nozzles. If not, explain why the nickel alloy welds of the reactor vessel nozzles are not identified as a limiting EAF location in LRA Table 4.3.4-2.

Luminant Response:

1. Table 4.3.4-2 of the LRA contains EAF results for the CPNPP piping locations. The CPNPP piping design consists solely of stainless steel; therefore, there are no carbon steel, low alloy steel, or nickel alloy locations in the current licensing basis fatigue evaluations (or evaluated for EAF in Table 4.3.4-2 of the LRA).

2. It was determined, in the CPNPP current licensing basis evaluations, that the nickel alloy material type for the reactor vessel nozzles is not fatigue sensitive (i.e. no usage values are included). The low-alloy material type reactor vessel inlet and outlet nozzle locations, defined in NUREG/CR-6260, are evaluated for EAF in Westinghouse report WCAP-18711-P (Proprietary) and WCAP-18711-NP (Non-Proprietary). Note, other degradation mechanisms (such as PWSCC, etc.) at the nickel alloy material type (weld) for the reactor vessel nozzles are addressed thru the PEO in other sections of the LRA (such as leak before break in LRA Section 4.7.1).

References:

None Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-3 TXX-23044 and CP-202300285 Enclosure 1.E Page 1 of 2 LRA Section:4.3.4, Environmentally Assisted Fatigue NRC RAI No: 4.3.4-3 (Accumulator Nozzle Crotch Region (180° Loc) 60-Yr Projected CUFen Discrepancy)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

The following reference indicates that there was a need to revise the CUFen calculations for the accumulator nozzles based on corrections to 60-year transient cycle projections

(

Reference:

Westinghouse CN-SDA-II-21-003-R0-ASMT-1, Assessment Record of CN-SDA-II-21-003 for CAP IR-2022-6325, August 18, 2022). The reference document also explains that the 60-year projection cycles of the following transients needed to be corrected: (1) refueling transient; (2) tube leak test transient; (3) accumulator line refueling transient; (4) reactor coolant system (RCS) venting transient; (5) reactor vessel stud tensioning transient; (6) accumulator check valve test transient.

The reference document further indicates that the projected cycles of the refueling transient and tube leak test transient are used in the EAF analysis for the accumulator nozzles.

Issue:

The revised 60-year projected CUFen value for the crotch region of the accumulator nozzle (180 degree location) in the reference is not consistent with that listed in LRA Table 4.3.4-2.

The nozzle crotch region is also called analysis section number 20. The reference document indicates that the 60-year projected CUFen value for the accumulator nozzle crotch region is 0.9941 in comparison with 0.974 listed in LRA Table 4.3.4-2.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-3 TXX-23044 and CP-202300285 Enclosure 1.E Page 2 of 2 Request:

1. Resolve the apparent inconsistency between the 60-year projected CUFen values of the accumulator nozzle crotch region listed in the reference document and LRA Table 4.3.4-2.

2. Clarify whether the corrected transient cycle projections need a revision to the other limiting locations and their CUFen values described in LRA Table 4.3.4-1 (equipment EAF) or 4.3.4-2 (piping EAF). If so, provide a revision to the tables.

3. Clarify whether the corrected transient cycle projections need a revision to LRA Table 4.3.1-2 that describes the 60-year projected cycles. If so, revise the table.

Luminant Response:

1. The fatigue usage value results documented in Westinghouse Proprietary document CN-SDA-II-21-003-R0-ASMT-1 are the current results. That is, the Accumulator Nozzle, 10 Accumulator Nozzle - Crotch Region entry in LRA Table 4.3.4-2 should include a CUFen value of 0.994 consistent with CN-SDA-II-21-003-R0-ASMT-1 in-place of the current CUFen value of 0.974.

2. It has been confirmed that the results, other than the accumulator location described in item 1, in LRA Table 4.3.4-1 (equipment EAF) or 4.3.4-2 (piping EAF) are accurate (and not impacted).

3. Similarly, it has been confirmed that the appropriate 60-year transient cycle projections are included in LRA Table 4.3.1-2.

References:

None Associated LRA Revisions:

LRA Section 4.3.4, Table 4.3.4-2 (page 4.3-20), is revised as follows:

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-4 TXX-23044 and CP-202300285 Enclosure 1.F Page 1 of 2 LRA Section: 4.3.4, Environmentally Assisted Fatigue NRC RAI No: 4.3.4-4 (TLAA Disposition Discrepancy)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

LRA Section 4.3.4 indicates that the environmentally assisted fatigue (EAF) TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(iii) and the aging effects of EAF will be managed by using the Fatigue Monitoring aging management program (AMP) and Steam Generators AMP. LRA Section A.3.3.4 describes the Final Safety Analysis Report (FSAR) supplement for the Fatigue Monitoring AMP.

Issue:

The FSAR supplement in LRA Section A.3.3.4 refers to 10 CFR 54.21(c)(1)(ii) instead of 10 CFR 54.21(c)(1)(iii) as a TLAA disposition, inconsistent with the TLAA disposition in LRA Section 4.3.4. Specifically, LRA Section A.3.3.4 states that the analyses have been projected to the end of the period of extended of operation and found that the CUF

[cumulative usage factor] will remain below the ASME Code allowable of 1.0 in accordance with 10 CFR 54.21(c)(1)(ii). The staff also needs clarification on why CUF instead of environmental CUF (CUFen) is referenced in the FSAR supplement sentence discussed above.

Request:

1. Provide justification for why the FSAR supplement for EAF TLAA refers to 10 CFR 54.21(c)(1)(ii), inconsistent with the TLAA disposition (i.e., 10 CFR 54.21(c)(1)(iii))

described in LRA Section 4.3.4. If justification cannot be provided, revise the FSAR supplement to delete the reference to 10 CFR 54.21(c)(1)(ii) as a TLAA disposition.

2. Explain why the CUF is mentioned rather than CUFen in the FSAR supplement sentence discussed in the issue section above.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.4-4 TXX-23044 and CP-202300285 Enclosure 1.F Page 2 of 2 Luminant Response:

1. The FSAR Supplement included in LRA Section A.3.3.4 incorrectly reflects a 10 CFR 54.21(c)(1)(ii) TLAA disposition which should be 10 CFR 54.21(c)(1)(iii) (consistent with LRA Section 4.3.4).

2. Similarly, use of the term "CUF" in the FSAR Supplement included in LRA Section A.3.3.4 is incorrect and should be "CUFen" (consistent with LRA Section 4.3.4).

References:

None Associated LRA Revisions:

LRA A.3.3.4, Paragraph 2 (page A-37), is revised as follows:

The environmental fatigue analyses prepared for the CPNPP Units 1 and 2 limiting components, equivalent to the locations evaluated in NUREG/CR-6260 for newer vintage Westinghouse plants, demonstrate that cumulative environmental fatigue usage values do not exceed the ASME allowable cumulative fatigue usage value of 1.0. Since the analyses are based on design cycles and 60-year cycle projections, monitoring of usage through the PEO is required to ensure these conclusions remain valid. Where reduced numbers of cycles were used in the environmental fatigue analyses, they will be considered the new CLB cycle limits in the Fatigue Monitoring Program during the PEO. The reduced numbers of cycles are equal to or greater than the 60-year projected cycles. Therefore, analyses have been projected to the end of the PEO and found that the CUFCUFen will remain below the ASME Code allowable of 1.0 in accordance with 10 CFR 54.21(c)(1)(iiiii).

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.5-1 TXX-23044 and CP-202300285 Enclosure 1.G Page 1 of 3 LRA Section: 4.3.5, Reactor Vessel Internals Fatigue Analyses NRC RAI No: 4.3.5-1 (Potential Inconsistency in Design Transients Between LRA Table and Reference)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

In its supplement dated April 6, 2023, the applicant revised LRA Table 4.3.1-2 with respect to the design cycles (40-year cycles) of the bypass line tempering valve transient.

Specifically, the applicant updated the design cycles of the transient from 20 cycles to 40 cycles.

Issue:

LRA Section 4.3.5 indicates that the following reference includes the most recent fatigue evaluations in the Comanche Peak Nuclear Power Plant (CPNPP) current licensing basis for reactor vessel internal (RVI) components (

Reference:

WCAP-16840-NP, Comanche Peak Nuclear Power Plant Stretch Power Uprate Licensing Report, Revision 0).

The staff noted a potential inconsistency in the design transients between LRA Table 4.3.1-2 and WCAP-16840-NP, Table 2.2.6-1. Specifically, LRA Table 4.3.1-2 includes the bypass line tempering valve transient, which is only applicable to CPNPP Unit 2. However, this transient is not included in WCAP-16840-NP, Table 2.2.6-1. In addition, WCAP-16840-NP, Table 2.2.6-1 includes the split flow bypass valve transient, which is only applicable to CPNPP Unit 2. However, this transient is not included in LRA Table 4.3.1-2.

The applicants supplement dated April 6, 2023 updates bypass line tempering valve transient from 20 cycles to 40 cycles. The staff needs to clarify (1) whether the updated design cycle number is consistent with the design cycle number in the current licensing basis and (2) the basis of the design cycle update.

The staff further noted that the 60-year projected cycles of the Unit 1 letdown flow shutoff with prompt return to service transient is greater than the design cycles in LRA Table 4.3.1-

3. The staff needs clarification on whether the transient cycles in LRA Table 4.3.1-3, including the Unit 1 letdown flow shutoff with prompt return to service transient cycles, are expected to increase the cumulative usage factors (CUFs) of RVI components above the design limit of 1.0 for the period of extended operation.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.5-1 TXX-23044 and CP-202300285 Enclosure 1.G Page 2 of 3 Request:

1. Clarify whether the bypass line tempering valve transient is identical to the split flow bypass valve transient. If not, resolve the inconsistency of these transients between LRA Table 4.3.1-2 and WCAP-16840-NP, Table 2.2.6-1.

2. Clarify (1) whether the updated design cycle number (40 cycles) of the bypass line tempering valve transient is consistent with the design cycles in the current licensing basis and (2) the basis of the design cycle update.

3. Clarify whether the transient cycles in LRA Table 4.3.1-3, including the Unit 1 letdown flow shutoff with prompt return to service transient cycles, are expected to increase the 60-year projected CUFs of RVI components above the design limit of 1.0. If so, discuss the affected components and their 60-year projected CUF values.

Luminant Response:

1. It has been confirmed that the LRA Table 4.3.1-2 "bypass line tempering valve" transient and the WCAP-16840-NP "split flow bypass valve" transient are the same transient. It should be noted that in note 4 of Table 2.2.6-1 of WCAP-16840-NP, it states that this transient, "split flow bypass valve", is included in the Unit 2 Model D-5 steam generator design specification. The Unit 2 Model D-5 steam generator design specification calls the same transient, "bypass line tempering valve failure."

2. 1) Additionally, it has been confirmed that the current licensing basis cycle limit for the "bypass line tempering valve" transient is 40 cycles; 20 cycles included in Table 4.3.1-2 of the LRA was incorrect.

2) The updated transient cycle limit, corrected by Reference 1, Attachment W4, is consistent with the current licensing basis uprate report WCAP-16840-NP, Table 2.2.6-1.

3. The 60-year transient cycle projections in Table 4.3.1-3 apply to the Class 1 auxiliary piping evaluations; therefore, there is no impact or increase in the 40-year usage values reported for the RVI (reactor vessel internals).

References:

1. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.5-1 TXX-23044 and CP-202300285 Enclosure 1.G Page 3 of 3 Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 1 of 8 LRA Section: Table 3.3.2-8c, Miscellaneous Ventilation Systems - Summary of Aging Management Evaluation NRC RAI No: 3.3.2.8c-1 (Aging Management of Aluminum Components Exposed to Outdoor Air)

Regulatory Basis:

10 CFR 54.21(a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a),

the staff requires additional information in regard to the matters described below.

Background:

LRA Table 3.3.2-8c, Miscellaneous Ventilation Systems - Summary of Aging Management Evaluation, states that aging effects for aluminum fan housings exposed to outdoor air are not applicable and no AMP is proposed. The AMR items cite generic note I and NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Revision 2, item 3.3.1-113.

As amended by letter dated April 24, 2023 (ML23114A377), SLRA Table 3.3-1, Summary of Aging Management Programs for Auxiliary Systems, item 113, states the following (in part):

[f]ans housings in the top floor of the EDG [emergency diesel generator] Building are also exposed to outdoor air. These fans remove heat from the EDG area through labyrinth missile protection features. Because the outdoor air environment at CPNPP [Comanche Peak Nuclear Power Plant] is non-aggressive, as described in Sections 3.3.2.2.3 and 3.3.2.2.5 [further evaluations for cracking and loss of material of stainless steel, respectively], the absence of aging effects listed in item 3.3-1, 113, for aluminum components exposed to indoor air, is also representative of equivalent components exposed to outdoor air where water cannot pool. NUREG-1800, Revision 2, item 3.3.1-81 states loss of material due to pitting and crevice corrosion is an applicable aging effect requiring management for aluminum piping, piping components, and piping elements exposed to outdoor air.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 2 of 8 Issue:

NUREG-1800, Revision 2, item 3.3.1-113 does not specifically address aging effects requiring management for aluminum components exposed to outdoor air (i.e., the item addresses only air-dry, air-indoor (uncontrolled and controlled), and gas environments).

However, the staffs position established in NUREG-1800, Revision 2, item 3.3.1-81 is that an outdoor air environment is sufficiently aggressive to result in loss of material due to pitting and crevice corrosion of aluminum components, irrespective of whether water is allowed to pool. In addition, the staffs position established in , Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, Section 3.3.2.2.10, Loss of Material Due to Pitting and Crevice Corrosion in Aluminum Alloys, (which can be considered as relevant operating experience during the first period of extended operation) is that moisture levels and halide concentrations should be considered high enough to facilitate pitting and/or crevice corrosion of aluminum alloys in atmospheric and uncontrolled air, unless demonstrated otherwise. Based on the above, the staff seeks further clarification with respect to why loss of material due to pitting and crevice corrosion is not aging effect requiring management for the aluminum fan housings exposed to outdoor air.

Request:

State the basis with respect to why loss of material due to pitting and crevice corrosion is not an applicable aging effect requiring management for the aluminum fan housings exposed to outdoor air. Alternatively, revise the LRA (as appropriate) to reflect that loss of material due to pitting and crevice corrosion will be managed for the subject components.

Luminant Response:

NUREG-1800, Revision 2, Section 3.2.2.2.3.2 addresses loss of material due to pitting and crevice corrosion for stainless steel piping, piping components, piping elements, and tanks exposed to outdoor air and states that pitting and crevice corrosion is only known to occur in environments containing sufficient halides (primarily chlorides) and in which condensation or deliquescence is possible. NUREG-2192 (which is considered as relevant operating experience during the first period of extended operation) contains a similar discussion of loss of material due to pitting and crevice corrosion in aluminum alloys and states that environments that can result in pitting and/or crevice corrosion of aluminum alloys are those that contain halides (e.g., chloride) in the presence of moisture.

Section 3.2.2.2.3 item 2 of the CPNPP LRA evaluates the outdoor air environment at CPNPP and concludes that the general ambient outdoor and indoor air is considered to be benign, non-aggressive lacking sufficient halides for corrosion of stainless steel. Based on the guidance in NUREG-2192 Section 3.2.2.2.10, which is considered as operating experience, the evaluation of the outdoor air environment is also applicable to corrosion of aluminum.

Furthermore, the aluminum fan housings are located in a sheltered location where they are not exposed to rainwater. A review of recent inspection results determined that visual inspections of the housings showed no evidence of degradation.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 3 of 8 Given that the outdoor air environment at CPNPP is considered to be benign, non-aggressive, and lacking sufficient halides for corrosion of aluminum and that the aluminum fan housings are sheltered from exposure to rainwater, loss of material due to pitting and crevice corrosion is not expected to occur in the aluminum fan housings exposed to outdoor air as supported by recent inspection results.

Although recent inspection results demonstrate that degradation has not been occurring in the aluminum fan housings, the supply air to the fans is outdoor air; therefore, loss of material due to crevice and pitting corrosion is conservatively included as an aging effect requiring management for the aluminum EDG ventilation fans exposed to outdoor air during the period of extended operation. LRA Table 3.3-1 (Items 3.3-1, 081 and 3.3-1, 113) and Table 3.3.2-8c are revised accordingly.

References:

None.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 4 of 8 Associated LRA Revisions:

LRA Table 3.3-1, Item 081 (page 3.3-60), is revised as follows:

Table 3.3-1 Summary of Aging Management Programs for Auxiliary Systems Item Number Component Aging Aging Management Further Evaluation Discussion Effect/Mechanism Programs Recommended 3.3-1, 081 Copper alloy, Aluminum Loss of material due Chapter XI.M36, No Consistent with NUREG-1801.

Piping, piping to pitting and External Surfaces The External Surfaces Monitoring of components, and piping crevice corrosion Monitoring of Mechanical Components (B.2.3.22) elements exposed to Mechanical AMP will manage loss of material for the Air - outdoor (External), Components aluminum insulation jacketing Air - outdoor commodity on components located outdoors, as listed in Table 3.5.2-13.

The copper alloy deluge spray nozzles exposed to outdoor air in the FPS are addressed in item 3.3-1, 131. There are no other copper alloy piping or piping components exposed to outdoor air in the auxiliary systems.

Aluminum fan housings exposed to outdoor air in the top floor of the EDG Building are conservatively assumed to be susceptible to loss of material due to pitting and crevice corrosion, despite the absence of this degradation mechanism throughout site OE.

Aluminum piping and ducting components exposed to indoor or outdoor air in the EDG and Auxiliary and Miscellaneous Ventilation Systems are addressed in item 3.3-1, 113. A generic note I and a plant-specific note are used for the outdoor air environment.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 5 of 8 LRA Table 3.3-1, Item 113 (page 3.3-77), is revised as follows:

Table 3.3-1 Summary of Aging Management Programs for Auxiliary Systems Item Number Component Aging Aging Management Further Evaluation Discussion Effect/Mechanism Programs Recommended 3.3-1, 113 Aluminum Piping, piping None None NA - No AEM or AMP Consistent with NUREG-1801.

components, and piping Aluminum fan housings and crankcase elements exposed to vacuum blowers exposed to indoor air in Air - dry the Miscellaneous Ventilation Systems (Internal/External), Air and EDG and Auxiliary Systems, indoor, uncontrolled respectively, are not subject to aging (Internal/External), Air effects.

indoor, controlled (External), Gas Fans housings in the top floor of the EDG Building are also exposed to outdoor air. These fans remove heat from the EDG area through labyrinth missile protection features. Because the outdoor air environment at CPNPP is non-aggressive, as described in Sections 3.3.2.2.3 and 3.3.2.2.5, the absence of aging effects listed in item 3.3-1, 113, for aluminum components exposed to indoor air, is also representative of equivalent components exposed to outdoor air where water cannot pool. A generic note I and plant-specific note are used.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 6 of 8 LRA Table 3.3.2-8c (page 3.3-240) is revised as follows:

Table 3.3.2-8c: Miscellaneous Ventilation Systems - Summary of Aging Management Evaluation Aging Effect Component Intended Aging Management NUREG-1801 Table 1 Notes Material Environment Requiring Type Function Programs Item Item Management Fan housing Pressure Aluminum Air - outdoor None None VII.J.AP-135 3.3-1, I, 9 boundary (external) Loss of material External Surfaces VII.I.AP-256 113 A Monitoring of Mechanical 3.3-1, Components (B.2.3.22) 081 Fan housing Pressure Aluminum Air - outdoor None None VII.J.AP-135 3.3-1, I, 9 boundary (internal) Loss of material External Surfaces VII.I.AP-256 113 A Monitoring of Mechanical 3.3-1, Components (B.2.3.22) 081

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 7 of 8 LRA Table 3.3.2-8c, Generic Notes (page 3.3-245), are revised as follows:

Generic Notes A. Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.

B. Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP takes some exceptions to NUREG-1801 AMP.

C. Component is different, but consistent with NUREG-1801 item for material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.

E. Consistent with NUREG-1801 item for material, environment, and aging effect, but a different AMP is credited or NUREG-1801 identifies a plant-specific AMP.

G. Environment not in NUREG-1801 for this component and material.

H. Aging effect not in NUREG-1801 for this component, material, and environment combination.

I. Aging effect in NUREG-1801 for this component, material and environment combination is not applicable.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.8c-1 TXX-23044 and CP-202300285 Enclosure 1.H Page 8 of 8 LRA Table 3.3.2-8c, Plant-Specific Note 9 (page 3.3-246), is revised as follows:

6. Consistent with the latest industry guidance, based on industry OE updates incorporated in NUREG-2191 (Item VII.F1.AP-103, Table 3.3-1, 096).

7. Consistent with the latest industry guidance, based on industry OE updates incorporated in NUREG-2191 (Item VII.I.A-720, Table 3.3-1, 150).

8. Consistent with the latest industry guidance, based on industry OE updates incorporated in NUREG-2191 (Item VII.I.A-716, Table 3.3-1, 151).

9. Aging effects not applicable, for more information see item Table 3.3-1, item 081. The absence of aging effects described in Table 3.3-1, item 113 is also representative for this component, material, and environment combination. Not used.

10. The Compressed Air Monitoring (B.2.3.14) AMP is applied to assure dry-air conditions are maintained during the PEO.

11. Insulated components associated with the UPS room A/C units operate below the dew point; therefore, these components have the potential to accumulate condensation which may trap particulates underneath insulation.

12. Stainless steel drip pans in the Miscellaneous Ventilation Systems are for the UPS room coolers.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.7.1-1 TXX-23044 and CP-202300285 Enclosure 1.I Page 1 of 2 LRA Section: 4.7.1, Leak-Before-Break NRC RAI No: 4.7.1-1 (Request for Submission of Supporting TLAA Document)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant shall demonstrate that (i) the analyses remain valid for the period of extended operation; (ii) the analyses have been projected to the end of the period of extended operation; or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

LRA Section 4.7.1 discusses the leak-before-break (LBB) TLAA for Comanche Peak Nuclear Power Plant (CPNPP). The applicant has dispositioned this TLAA in accordance with 10 CFR 54.21(c)(1)(ii).

Issue:

During the audit, the applicant indicated that Westinghouse WCAP-10527 Revision 3, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Comanche Peak Units 1 and 2 for the License Renewal Program (60 Years), January 2023 (Proprietary) and Westinghouse Report WCAP-10528, Revision 3, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Comanche Peak Units 1 and 2 for the License Renewal Program (60 Years), January 2023 (Non-Proprietary) contain the updated LBB evaluations for CPNPP Unit 1 and Unit 2 assuming a 60-year plant life.

The applicant indicated that the LBB evaluations for CPNPP Unit 1 and Unit 2 were updated to account for implementation and partial implementation of the Material Stress Improvement Process (MSIP). The applicant also explained that dissimilar metal weld locations at RPV nozzles, which have Alloy 82/182 nickel-base materials and are susceptible to primary water stress corrosion cracking (PWSCC), were evaluated to confirm that those locations have been appropriately evaluated for LBB. The applicant further explained that the Alloy 82/182 welds have been conservatively evaluated to consider the effects of PWSCC. However, the applicant did not submit the updated LBB evaluations for NRC staffs review as part of the LRA.

Request:

Since the LBB evaluations have been updated for the period of extended operation (PEO),

please submit WCAP-10527 Revision 3 (Proprietary) and WCAP-10528, Revision 3, (Non-proprietary) for NRC staffs review on the docket.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.7.1-1 TXX-23044 and CP-202300285 Enclosure 1.I Page 2 of 2 Luminant Response:

As requested, Westinghouse has provided WCAP-10527 Revision 3 (Proprietary) within Enclosure 3 and WCAP-10528, Revision 3, (Non-proprietary) within Enclosure 2.

References:

None Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-1 TXX-23044 and CP-202300285 Enclosure 1.J Page 1 of 2 LRA Section: B.2.3.3, Reactor Head Closure Stud Bolting NRC RAI No: RAI B.2.3.3-1 (Reactor Head Closure Stud Thread Damage)

Regulatory Basis:

Title 10 of the Code of Federal Regulations (CFR) Section 54.21(a)(1) requires license renewal applicants to perform an integrated plant assessment in their application to identify and list systems, structures, and components (SSCs) that are within the scope of license renewal and subject to aging management review (AMR). Further, 10 CFR 54.21(a)(3) requires that for the SSCs identified to be subject to AMR, the applicants demonstrate that the effects of aging will be adequately managed such that their intended functions are maintained consistent with the current licensing basis for the period of extended operation (PEO). To complete its review and enable the staff to make a reasonable assurance finding regarding the functionality of the reviewed SSCs for the PEO, the staff needs additional information regarding the matters described below. This information is described in the following requests for additional information (RAIs).

Background:

Comanche Peak Nuclear Power Plant (CPNPP) LRA Section B.2.3.3 describes the applicants aging management program (AMP) for the reactor head closure stud bolting (studs, nuts, washers, and threads-in-flange) of the CPNPP units. In the plant-specific operating experience section of the subject AMP, the applicant described thread damage that occurred over time in the studs and threads-in-flange (i.e., the stud holes). This RAI is to obtain information on the number of stud locations that were damaged.

Issue:

LRA Section B.2.3.3 did not state how many stud locations were damaged nor if there were recent or new thread damage since the evaluation of the damage in 2014.

Request:

a. State how many stud locations were damaged at each unit.

b. State whether the number of damaged threads increased for any stud or stud hole that had previously damaged threads, especially the one stud hole with 13.75 missing threads since 2014.

c. State whether there has been any recent or new thread damage in any stud or stud hole since 2014.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-1 TXX-23044 and CP-202300285 Enclosure 1.J Page 2 of 2 Luminant Response:

Note: For the purposes of this response the word damaged refers to threads (or portions of threads) which have been entirely removed from a stud or associated stud hole in the reactor vessel flange. Partially degraded threads which are lapped, refiled, or otherwise reworked and returned to service are not included within the population of damaged studs or stud holes.

a. CPNPP Units 1 and 2 each utilize 54 reactor head closure studs to tension their respective reactor vessel heads. For Unit 1, 4 of the 54 stud locations include damaged threads (locations 7, 11, 17, and 25). For Unit 2, 9 of the 54 stud locations include damaged threads (locations 3, 5, 6, 12, 20, 30, 44, 45, and 54). Damage in these locations, including the location with 13.75 missing threads, is inclusive of threads on both the stud and the associated stud hole.

b. Of the damaged stud locations described in part (a) above, none of these studs or stud holes have experienced further damage since 2014. Furthermore, existing thread damage at all of these aforementioned locations was found to have occurred early in the life of the plant, around the 1991 to 1996 timeframe.

c. As alluded to in part (b), new thread damage has not occurred in any stud or stud hole since 2014, and existing thread damage can be attributed to events early in the operating life of the two units.

References:

None.

Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-2 TXX-23044 and CP-202300285 Enclosure 1.K Page 1 of 2 LRA Section: B.2.3.3, Reactor Head Closure Stud Bolting NRC RAI No: RAI B.2.3.3-2 (Reactor Head Closure Stud Thread Damage Corrective Actions)

Regulatory Basis:

Title 10 of the Code of Federal Regulations (CFR) Section 54.21(a)(1) requires license renewal applicants to perform an integrated plant assessment in their application to identify and list systems, structures, and components (SSCs) that are within the scope of license renewal and subject to aging management review (AMR). Further, 10 CFR 54.21(a)(3) requires that for the SSCs identified to be subject to AMR, the applicants demonstrate that the effects of aging will be adequately managed such that their intended functions are maintained consistent with the current licensing basis for the period of extended operation (PEO). To complete its review and enable the staff to make a reasonable assurance finding regarding the functionality of the reviewed SSCs for the PEO, the staff needs additional information regarding the matters described below. This information is described in the following requests for additional information (RAIs).

Background:

In the plant-specific operating experience section of the subject AMP, the applicant described thread damage that occurred over time in the studs and threads-in-flange (i.e., the stud holes). This RAI is to obtain information on the factors contributing to the thread damage and corrective actions taken when thread damage is found.

Issue:

LRA Section B.2.3.3 did not state the factors contributing to the subject thread damage nor the corrective actions taken when thread damage is found.

Request:

a. Describe the factors that contributed to the thread damage in the studs and stud holes.

b. Describe the corrective actions (e.g., use of lubricants, lowering tensioning limits) per 10 CFR 50 Appendix B via the Corrective Actions program element of the subject AMP, to prevent such damage in the future.

c. Describe whether the corrective actions on the damaged studs/stud holes include repair and/or replacement.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-2 TXX-23044 and CP-202300285 Enclosure 1.K Page 2 of 2 Luminant Response:

Note: For the purposes of this response the word damaged refers to threads (or portions of threads) which have been entirely removed from a stud or associated stud hole in the reactor vessel flange. Partially degraded threads which are lapped, refiled, or otherwise reworked and returned to service are not included within the population of damaged studs or stud holes.

a. The primary contributor to the thread damage in the studs and stud holes was galling of the threads, which can be induced by factors such as improper lubrication.

b. Corrective actions in response to degraded thread events include reworking affected locations (i.e., lapping or refiling). Those threads which cannot be reworked are removed.

Furthermore, proper and adequate use of station-approved lubricants and adherence to procedurally driven hydraulic tensioning limits aides in the prevention of future thread degradation.

Additionally, any future deficient conditions, should they occur, will be evaluated, dispositioned, and corrected within the Corrective Action Program, in accordance with 10 CFR 50 Appendix B.

c. As stated in the response to part (b), corrective actions may include reworking of degraded threads or thread removal. Additionally, for those locations with damaged threads, should the number of missing threads for a given stud and stud hole location approach the maximum allowable limit, further replacement or repair activities are evaluated.

Replacement activities include use of replacement studs and repair activities include sleeving of stud holes within the reactor vessel flange.

References:

None.

Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-3 TXX-23044 and CP-202300285 Enclosure 1.L Page 1 of 2 LRA Section: B.2.3.3, Reactor Head Closure Stud Bolting NRC RAI No: RAI B.2.3.3-3 (Reactor Head Closure Stud Tension)

Regulatory Basis:

Title 10 of the Code of Federal Regulations (CFR) Section 54.21(a)(1) requires license renewal applicants to perform an integrated plant assessment in their application to identify and list systems, structures, and components (SSCs) that are within the scope of license renewal and subject to aging management review (AMR). Further, 10 CFR 54.21(a)(3) requires that for the SSCs identified to be subject to AMR, the applicants demonstrate that the effects of aging will be adequately managed such that their intended functions are maintained consistent with the current licensing basis for the period of extended operation (PEO). To complete its review and enable the staff to make a reasonable assurance finding regarding the functionality of the reviewed SSCs for the PEO, the staff needs additional information regarding the matters described below. This information is described in the following requests for additional information (RAIs).

Background:

In the plant-specific operating experience section of the subject AMP, the applicant described thread damage that occurred over time in the studs and threads-in-flange (i.e., the stud holes). This RAI is to clarify information on the allowable number of missing threads.

Issue:

LRA Section B.2.3.3 stated that the maximum allowable number of missing threads is 17.22, which was previously 13.1. Based on its audit of the AMP, the staff noted that for the previous maximum allowable number of missing threads of 13.1 there was no limit on stud tension/detension cycles. However, it is not clear to the staff whether there is a limit on stud tension/detension cycles for the new limit of 17.22 missing threads.

Request:

a. Clarify whether there is a limit on stud tension/detension cycles for the new limit of 17.22 missing threads.

b. If there is a limit, state the value of this limit on the number of stud tension/detension cycles and explain how this limit will be met to the end of the PEO.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-3 TXX-23044 and CP-202300285 Enclosure 1.L Page 2 of 2 Luminant Response:

a. No, there is not a limit on stud tension/detension cycles for the updated limit of 17.22 missing threads. The calculation which derives the updated allowable missing thread limit of 17.22 threads is based on the maximum tensioning load experienced by the studs. The maximum tensioning load condition is a criterion which is not based on cyclic loading or fatigue evaluation. Furthermore, the stud tensioning procedures for each unit include disclaimers which prohibit over-tensioning of the studs past this maximum tensioning condition (i.e., 9300 psi or 2350 kips).

b. As stated in the response to part (a), the maximum allowable missing thread limit of 17.22 threads is not dependent on cyclic loading or fatigue evaluation; therefore, the maximum allowable missing thread limit is not contingent upon tension/detension cycles. As such, there is no limit on the number of stud tension/detension cycles through the end of the PEO.

References:

None.

Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-4 TXX-23044 and CP-202300285 Enclosure 1.M Page 1 of 2 LRA Section: B.2.3.3, Reactor Head Closure Stud Bolting NRC RAI No: RAI B.2.3.3-4 (Reactor Head Closure Stud AMP Enhancements)

Regulatory Basis:

Title 10 of the Code of Federal Regulations (CFR) Section 54.21(a)(1) requires license renewal applicants to perform an integrated plant assessment in their application to identify and list systems, structures, and components (SSCs) that are within the scope of license renewal and subject to aging management review (AMR). Further, 10 CFR 54.21(a)(3) requires that for the SSCs identified to be subject to AMR, the applicants demonstrate that the effects of aging will be adequately managed such that their intended functions are maintained consistent with the current licensing basis for the period of extended operation (PEO). To complete its review and enable the staff to make a reasonable assurance finding regarding the functionality of the reviewed SSCs for the PEO, the staff needs additional information regarding the matters described below. This information is described in the following requests for additional information (RAIs).

Background:

LRA Section B.2.3.3 listed enhancements to certain program elements of the subject AMP.

Issue:

Based on the staffs audit of the AMP program basis document, the staff noted that one of the enhancements in LRA Section B.2.3.3 is inconsistent with the AMP program basis document in that the enhancement regarding the revision of the procurement requirements to assure the proper yield strength for replacement materials is associated with the Scope of Program program element instead of the Preventive Actions program element.

Request:

Clarify whether the enhancement regarding the revision of the procurement requirements to assure the proper yield strength for replacement materials should be associated with the Preventive Actions program element of the AMP (as it is in the program basis document),

not with the Scope of Program program element of the AMP.

Luminant Response:

As described in the Reactor Head Closure Stud Bolting AMP program basis document, the enhancement regarding the revision of the procurement requirements to assure the proper yield strength for replacement materials should be associated with only the Preventive Actions and Corrective Actions program elements. A revision to LRA Section B.2.3.3 is included below for clarity.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.3-4 TXX-23044 and CP-202300285 Enclosure 1.M Page 2 of 2

References:

None.

Associated LRA Revisions:

LRA Section B.2.3.3, Enhancements (page B-39), is revised as follows:

Enhancements The Reactor Head Closure Stud Bolting AMP will be enhanced as follows for alignment with NUREG-1801,Section XI.M3. The changes and enhancements are to be implemented no later than six months prior to entering the PEO or no later than the last refueling outage prior to the PEO.

Element Affected Enhancement

1. Scope Revise the procurement requirements

2. Preventive Actions and/or engineering specification for the reactor head closure stud bolting material to

7. Corrective Actions assure the maximum yield strength of replacement reactor head closure stud material purchased in the future is limited to a measured yield strength of <150 ksi.

2. Preventive Actions Revise maintenance documents for the installation of the reactor vessel head to explicitly prohibit the use of lubricants not meeting RG 1.65 guidance.

Enclosure 2:

Westinghouse Report WCAP-10528, Revision 3, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Comanche Peak Units 1 and 2 for the License Renewal Program (60 Years)." -

Non-Proprietary 77 pages follow

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-10528-NP January 2023 Revision 3 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Comanche Peak Units 1 and 2 for the License Renewal Program (60 Years)

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ii WCAP-10528-NP Revision 3 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Comanche Peak Units 1 and 2 for the License Renewal Program (60 Years)

January 2023 Author: Kevin Weldon*

Structural Design and Analysis Timothy J. Nowicki*

Operating Plants Piping and Supports Reviewer: Momo Wiratmo*

Operating Plants Piping and Supports Approved: Lynn A. Patterson, Manager*

Reactor Vessel and Containment Vessel Design and Analysis

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

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

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii REVISION INDEX Revision Date Remarks 1 See Prime Non-Proprietary Version of WCAP-10527-P Revision 1.

Non-Proprietary Version of WCAP-10527-P Revision 2 Revised to include the potential of a partial MSIP configuration for Unit 2.

Added specific evaluations and conclusions for the alloy 82/182 material at the reactor pressure vessel (RPV) nozzle.

2 See Prime Update to include March 2021 Errata for NUREG/CR-4513 Revision 2.

Minor Formatting Updates.

Changes are shown via revision bars.

Revised Table 8-3 to include updated 60-year transient cycle projections found in Table 2 of LTR-SDA-II-21-28 Revision 2, which summarizes the CPNPP FSAR 60-year transient cycle projections. LTR-SDA-II-21-28 3 See PRIME Revision 2 has been added to this document as Reference 8-8.

Changes are shown via revision bars.

WCAP-10528-NP January 2023 Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS LIST OF TABLES ............................................................................................................................................. v LIST OF FIGURES ......................................................................................................................................... vii 1 Background and Purpose ...................................................................................................... 1-1

1.1 BACKGROUND

INFORMATION ......................................................................... 1-1 1.2 SCOPE AND OBJECTIVES ................................................................................... 1-2

1.3 REFERENCES

........................................................................................................ 1-4 2 Operation and Stability of the Reactor Coolant System ....................................................... 2-1 2.1 STRESS CORROSION CRACKING ..................................................................... 2-1 2.2 WATER HAMMER ................................................................................................. 2-2 2.3 LOW CYCLE AND HIGH CYCLE FATIGUE ...................................................... 2-3 2.4 WALL THINNING, CREEP, AND CLEAVAGE .................................................... 2-3

2.5 REFERENCES

........................................................................................................ 2-3 3 Pipe Geometry and Loading ................................................................................................. 3-1

3.1 INTRODUCTION

TO METHODOLOGY ............................................................. 3-1 3.2 CALCULATION OF LOADS AND STRESSES .................................................... 3-2 3.3 LOADS FOR LEAK RATE EVALUATION ........................................................... 3-2 3.4 LOAD COMBINATION FOR CRACK STABILITY ANALYSES........................ 3-3

3.5 REFERENCES

........................................................................................................ 3-3 4 Material Characterization...................................................................................................... 4-1 4.1 PRIMARY LOOP PIPE AND FITTINGS MATERIALS........................................ 4-1 4.2 TENSILE PROPERTIES ......................................................................................... 4-1 4.3 FRACTURE TOUGHNESS PROPERTIES ............................................................ 4-2

4.4 REFERENCES

........................................................................................................ 4-5 5 Critical Location and Evaluation Criteria ............................................................................. 5-1 5.1 CRITICAL LOCATIONS ........................................................................................ 5-1 5.2 EVALUATION CRITERIA ..................................................................................... 5-1 6 Leak Rate Predictions ........................................................................................................... 6-1

6.1 INTRODUCTION

................................................................................................... 6-1 6.2 GENERAL CONSIDERATIONS ............................................................................ 6-1 6.3 CALCULATION METHOD ................................................................................... 6-1 6.4 LEAK RATE CALCULATIONS............................................................................. 6-2

6.5 REFERENCES

........................................................................................................ 6-2 7 Fracture Mechanics Evaluation............................................................................................. 7-1 7.1 LOCAL FAILURE MECHANISM ......................................................................... 7-1 7.2 GLOBAL FAILURE MECHANISM ...................................................................... 7-2

7.3 REFERENCES

........................................................................................................ 7-3 8 Fatigue Crack Growth Analysis ............................................................................................ 8-1

8.1 REFERENCES

........................................................................................................ 8-3 9 Assessment of Margins ......................................................................................................... 9-1 10 Conclusions......................................................................................................................... 10-1 APPENDIX A Limit Moment...................................................................................................... A-1 WCAP-10528-NP January 2023 Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 v LIST OF TABLES Table 3-1 Dimensions, Normal Loads and Stresses for Comanche Peak Unit 1 ....................................... 3-4 Table 3-2A Dimensions, Normal Loads and Stresses for Comanche Peak Unit 2 (without MSIP)........... 3-5 Table 3-2B Dimensions, Normal Loads and Stresses for Comanche Peak Unit 2 (with MSIP)................ 3-6 Table 3-3 Dimensions, Faulted Loads and Stresses for Comanche Peak Unit 1 ....................................... 3-7 Table 3-4A Dimensions, Faulted Loads and Stresses for Comanche Peak Unit 2 (without MSIP) ........... 3-8 Table 3-4B Dimensions, Faulted Loads and Stresses for Comanche Peak Unit 2 (with MSIP) ................ 3-9 Table 4-1 Measured Tensile Properties for Comanche Peak Unit 1 Primary Loop Pipes and Elbows (A351-CF8A) .............................................................................................................................. 4-6 Table 4-2 Measured Tensile Properties for Comanche Peak Unit 2 RCL Pipes and Elbows (A351-CF8A)4-7 Table 4-3 ASME Code Tensile Properties for Material A351-CF8A ......................................................... 4-9 Table 4-4 Tensile Properties for Comanche Peak Units 1 and 2 A351-CF8A Materials at Operating Temperatures .................................................................................................................. 4-10 Table 4-5 Comanche Peak Unit 1 CF8A Chemical Composition ............................................................ 4-11 Table 4-6 Comanche Peak Unit 2 CF8A Chemical Composition ............................................................ 4-12 Table 4-7 Comanche Peak Unit 1 CF8A Fracture Toughness and Tearing Modulus Properties .............. 4-13 Table 4-8 Comanche Peak Unit 2 CF8A Fracture Toughness and Tearing Modulus Properties .............. 4-14 Table 4-9 Comanche Peak Units 1 and 2 CF8A with Lowest Fracture Toughness Properties................ 4-15 Table 4-10 JIC Limiting Values NUREG/CR-4513 either per Rev. 1 or per Rev. 2 for Unit 1 ................ 4-16 Table 4-11 JIC Limiting Values NUREG/CR-4513 either per Rev. 1 or per Rev. 2 for Unit 2 ................ 4-17 Table 4-12 Alloy 82/182 Tensile Properties ............................................................................................. 4-18 Table 5-1 Critical Analysis Locations for Comanche Peak Units 1 and 2 RCL Lines ............................... 5-2 Table 6-1 Flaw Sizes for Comanche Peak Units 1 and 2 Yielding a Leak Rate of 10 gpm for the RCL Lines with A351-CF8A Material ............................................................................................... 6-3 Table 6-2 Flaw Sizes for Comanche Peak Units 1 and 2 Yielding a Leak Rate of 10 gpm for the Critical Analysis Location with Alloy 82/182 Welds ................................................................... 6-4 Table 7-1 Stability Results for Comanche Peak Units 1 and 2 Based on J-Integral Evaluations and CASS Thermal Aging Material................................................................................................... 7-4 Table 7-2 Flaw Stability Results for Comanche Peak Units 1 and 2 RCL Lines Based on Limit Load with A351-CF8A Material ....................................................................................................... 7-5 Table 7-3 Flaw Stability Results for Comanche Peak Units 1 and 2 RCL Lines Based on Limit Load with Alloy 82/182 welds .......................................................................................................... 7-6 Table 8-1 Summary of Transients (Representative 60-Year Design) ......................................................... 8-4 WCAP-10528-NP January 2023 Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi Table 8-2 Typical Fatigue Crack Growth at [ ] a,c,e (60 Years) .......................... 8-5 Table 8-3 Summary of Transients for Comanche Peak Units 1 and 2 ....................................................... 8-6 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Comanche Peak Units 1 and 2 with A351-CF8A CASS Material ...................................................................................................... 9-1 Table 9-2 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Comanche Peak Units 1 and 2 with Alloy 82/182 Welds ................................................................................................................... 9-2 WCAP-10528-NP January 2023 Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii LIST OF FIGURES Figure 3-1 Hot Leg Coolant Pipe ................................................................................................... 3-10 Figure 3-2 Comanche Peak Units 1 and 2 RCL Weld Locations ..................................................... 3-11 Figure 4-1 Pre-Service J vs. a for Cast Stainless Steel at 600F ................................................... 4-19 Figure 5-1 Schematic Diagram of Comanche Peak Units 1 and 2 Primary Loop Showing Critical Weld Locations ................................................................................................... 5-3 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures ....................... 6-5 Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of L/D .......................................... 6-6 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack .......................................... 6-7 Figure 7-1 [ ]a,c,e Stress Distribution ............................................................................ 7-7 Figure 8-1 Typical Cross-Section of [ ]a,c,e ............................................ 8-8 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels ....... 8-9 Figure 8-3 Reference Fatigue Crack Growth Law for [ ]a,c,e in a Water Environment at 600°F .................................................................................................... 8-10 Figure A-1 Pipe with a Through-Wall Crack in Bending ................................................................ A-2 WCAP-10528-NP January 2023 Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 1 BACKGROUND AND PURPOSE This report applies to the Comanche Peak Units 1 and 2 Reactor Coolant System (RCS) primary loop piping.

It is intended to demonstrate that for the specific parameters of the Comanche Peak Units 1 and 2 Nuclear Power Plants, RCS primary loop pipe breaks need not be considered in the structural design basis for the 60-year plant life License Renewal (LR) program. The specific parameters include normal operation temperature based on stretch power uprate programs and also NSSS design transient cycles for 60-year plant life.

Originally, Westinghouse performed the Leak-Before-Break (LBB) evaluation in 1984 for the Comanche Peak Units 1 and 2 primary loop piping (Reference 1-7). Later in 1989, the LBB loads were revised due to Units 1 and 2 primary loop piping Steam Generator (SG) snubber optimization, revised concrete and support stiffness. In 1998, the LBB evaluation was updated for 4.5% power uprating program for both Units 1 and 2. The revised loads that were examined against the loads used in Reference 1-7 and it was concluded that Reference 1-7 evaluation conclusions remained applicable.

The LBB evaluation for Unit 1 was updated in 2005 due to Replacement Steam Generators (RSG) / uprating and SG snubber elimination programs.

In 2007, the LBB loads were reviewed for stretch uprating program implemented for both Unit 1 and Unit

2. In 2019, Unit 1 LBB evaluation was updated due to Mechanical Stress Improvement Process (MSIP) implementation for Comanche Peak Unit 1 reactor pressure vessel (RPV) nozzles. In 2021, Unit 2 LBB evaluation was updated due to MSIP implementation for Unit 2 RPV nozzles. Because the License Renewal application may be submitted prior to implementing MSIP, this report will address Unit 2 evaluations and results both with and without MSIP.

For the 60-year plant life LR program, this report also reviews the dissimilar metal weld (DMW) locations at the Units 1 and 2 RPV nozzles that have Alloy 82/182 nickel-base materials which are susceptible to primary water stress corrosion cracking (PWSCC), to confirm that those locations have been appropriately mitigated by the application of MSIP and evaluated for LBB. In addition to RPV inlet and outlet nozzles, other locations susceptible for PWSCC such as the pressurizer safety and relief, spray and surge nozzles have also been repaired with the application of Structural Weld Overlay (SWOL).

1.1 BACKGROUND

INFORMATION Westinghouse has performed considerable testing and analysis to demonstrate that Reactor Coolant System (RCS) primary loop pipe breaks can be eliminated from the structural design basis of all Westinghouse plants. The concept of eliminating pipe breaks in the RCS primary loop was first presented to the NRC in 1978 in WCAP-9283 (Reference 1-1). That topical report employed a deterministic fracture mechanics evaluation and a probabilistic analysis to support the elimination of RCS primary loop pipe breaks. That approach was then used as a means of addressing Generic Issue A-2 and Asymmetric Loss of Coolant Accident (LOCA) Loads.

Westinghouse performed additional testing and analysis to justify the elimination of RCS primary loop pipe breaks. This material was provided to the NRC along with Letter Report NS-EPR-2519 (Reference 1-2).

Background and Purpose January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 The NRC funded research through Lawrence Livermore National Laboratory (LLNL) to address this same issue using a probabilistic approach. As part of the LLNL research effort, Westinghouse performed extensive evaluations of specific plant loads, material properties, transients, and system geometries to demonstrate that the analysis and testing previously performed by Westinghouse and the research performed by LLNL applied to all Westinghouse plants (References 1-3 and 1-4). The results from the LLNL study were released at a March 28, 1983, ACRS Subcommittee meeting. These studies, which are applicable to all Westinghouse plants east of the Rocky Mountains, determined the mean probability of a direct LOCA (RCS primary loop pipe break) to be 4.4 x 10-12 per reactor year and the mean probability of an indirect LOCA to be 10-7 per reactor year. Thus, the results previously obtained by Westinghouse (Reference 1-1) were confirmed by an independent NRC research study.

Based on the studies by Westinghouse, LLNL, the ACRS, and the Atomic Industrial Forum (AIF), the NRC completed a safety review of the Westinghouse reports submitted to address asymmetric blowdown loads that result from a number of discrete break locations on the PWR primary systems. The NRC Staff evaluation (Reference 1-5) concludes that an acceptable technical basis has been provided so that asymmetric blowdown loads need not be considered for those plants that can demonstrate the applicability of the modeling and conclusions contained in the Westinghouse response or can provide an equivalent fracture mechanics demonstration of the primary coolant loop integrity. In a more formal recognition of Leak-Before-Break (LBB) methodology applicability for PWRs, the NRC appropriately modified 10 CFR 50, General Design Criterion 4, Requirements for Protection Against Dynamic Effects for Postulated Pipe Rupture (Reference 1-6).

For Comanche Peak Units 1 and 2, the postulated pipe breaks for the RCS primary loop piping have been evaluated using LBB evaluation methods. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the primary loop piping need not be considered in the structural design basis of Comanche Peak Units 1 and 2. The original LBB evaluation results for the RCS primary loop were documented in WCAP-10527 report (Reference 1-7) in 1984.

For the License Renewal (LR) program, this report demonstrates that the conclusions reached in Reference 1-7 remain applicable in the structural design basis for the 60-year plant life for the specific parameters of the Comanche Peak Units 1 and 2.

1.2 SCOPE AND OBJECTIVES The general purpose of this investigation is to demonstrate leak-before-break for the primary loops in Comanche Peak Units 1 and 2 on a plant specific basis for the 60-year plant life. The recommendations and criteria proposed in References 1-8 and 1-9 are used in this evaluation. These criteria and resulting steps of the evaluation procedure can be briefly summarized as follows:

1. Calculate the applied loads. Identify the locations at which the highest stress occurs.

2. Identify the limiting material profiles and the associated material properties.

3. Postulate a surface flaw at the governing locations. Determine fatigue crack growth.

Show that a through-wall crack will not result.

4. Postulate a through-wall flaw at the governing locations. The size of the flaw should be large enough so that the leakage is assured of detection with margin using the installed leak detection Background and Purpose January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-3 equipment when the pipe is subjected to normal operating loads. A margin of 10 is demonstrated between the calculated leak rate and the leak detection capability.

5. Using faulted loads, demonstrate that there is a margin of 2 between the leakage flaw size and the critical flaw size.

6. Review the operating history to ascertain that operating experience has indicated no particular susceptibility to failure from the effects of corrosion, water hammer, fatigue, or other potential degradation mechanisms.

7. For the materials actually used in the plant provide the properties including toughness and tensile test data. Evaluate long term effects such as thermal aging.

8. Demonstrate margin on the calculated applied load value; margin of 1.4 using algebraic summation of loads or margin of 1.0 using absolute summation of loads.

This report provides a fracture mechanics demonstration of primary loop integrity for the Comanche Peak Units 1 and 2 plants consistent with the NRC position for exemption from consideration of dynamic effects.

The LBB evaluation summarized in this report consider the limiting weld locations of the RCL piping. In general, the analyses consider the material properties of the piping base metal, which are more limiting than the weld materials. The re-evaluations were performed to ensure that the LBB evaluation conclusions remain valid for 60-year plant life in the LR program.

It should be noted that the terms flaw and crack have the same meaning and are used interchangeably.

Governing location and critical location are also used interchangeably throughout the report.

The computer codes used in this evaluation for leak rate and fracture mechanics calculations have been validated and used for all the LBB applications by Westinghouse.

Note that there are several locations in this report where proprietary information has been identified and bracketed. For each of the bracketed locations, the reason for the proprietary classification is given using a standardized system. The proprietary brackets are labeled with three different letters, to provide this information, and the explanation for each letter is given below:

a. The information reveals the distinguishing aspects of a process or component, structure, tool, method, etc., and the prevention of its use by Westinghouses competitors, without license from Westinghouse, gives Westinghouse a competitive economic advantage.

c. The information, if used by a competitor, would reduce the competitors expenditure of resources or improve the competitors advantage in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

e. The information reveals aspects of past, present, or future Westinghouse or customer-funded development plans and programs of potential commercial value to Westinghouse.

The proprietary information in the brackets which have been deleted in this version of this report are provided in the proprietary class 2 document (WCAP-10527-P, Rev 2).

Background and Purpose January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-4

1.3 REFERENCES

1-1 WCAP-9283, The Integrity of the Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events, March 1978.

1-2 Westinghouse Proprietary Class 2 Letter Report NS-EPR-2519, Westinghouse (E. P. Rahe) to NRC (D. G. Eisenhut), November 10, 1981.

1-3 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston), April 25, 1983.

1-4 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston), July 25, 1983.

1-5 USNRC Generic Letter 84-04,

Subject:

Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops, February 1, 1984.

1-6 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

1-7 WCAP-10527, Revision 0, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Comanche Peaks Units 1 and 2, April 1984.

1-8 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday August 28, 1987/Notices, pp. 32626-32633.

1-9 NUREG-0800 Revision 1, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures, March 2007.

Background and Purpose January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2 OPERATION AND STABILITY OF THE REACTOR COOLANT SYSTEM 2.1 STRESS CORROSION CRACKING The Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking (IGSCC)).

This operating history totals over 1400 reactor-years, including 16 plants each having over 30 years of operation, 10 other plants each with over 25 years of operation, 11 plants each w i t h over 20 years of operation, and 12 plants each w i t h over 15 years of operation.

In 1978, the United States Nuclear Regulatory Commission (USNRC) formed the second Pipe Crack Study Group. (The first Pipe Crack Study Group (PCSG) established in 1975, addressed cracking in boiling water reactors only.) One of the objectives of the second PCSG was to include a review of the potential for stress corrosion cracking in Pressurized Water Reactors (PWRs). The results of the study performed by the PCSG were presented in NUREG-0531 (Reference 2-1) entitled Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants. In that report the PCSG stated:

The PCSG has determined that the potential for stress-corrosion cracking in PWR primary system piping is extremely low because the ingredients that produce IGSCC are not all present. The use of hydrazine additives and a hydrogen overpressure limit the oxygen in the coolant to very low levels. Other impurities that might cause stress-corrosion cracking, such as halides or caustic, are also rigidly controlled. Only for brief periods during reactor shutdown when the coolant is exposed to the air and during the subsequent startup are conditions even marginally capable of producing stress-corrosion cracking in the primary systems of PWRs.

Operating experience in PWRs supports this determination. To date, no stress corrosion cracking has been reported in the primary piping or safe ends of any PWR.

During 1979, several instances of cracking in PWR feedwater piping led to the establishment of the third PCSG. The investigations of the PCSG reported in NUREG-0691 (Reference 2-2) further confirmed that no occurrences of IGSCC have been reported for PWR primary coolant systems.

As stated above, for the Westinghouse plants there is no history of cracking failure in the reactor coolant system loop. The discussion below further qualifies the PCSGs findings.

For stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, susceptible material, and a corrosive environment. Since some residual stresses and some degree of material susceptibility exist in any stainless steel piping, the potential for stress corrosion is minimized by properly selecting a material immune to SCC as well as preventing the occurrence of a corrosive environment. The material specifications consider compatibility with the systems operating environment (both internal and external) as well as other material in the system, applicable ASME Code rules, fracture toughness, welding, fabrication, and processing.

The elements of a water environment known to increase the susceptibility of austenitic stainless steel to stress corrosion are: oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g., sulfides, sulfites, and thionates). Strict pipe cleaning standards prior to operation Operation and Stability of the Reactor Coolant System January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally.

During flushes and preoperational testing, water chemistry is controlled in accordance with written specifications. Requirements on chlorides, fluorides, conductivity, and pH are included in the acceptance criteria for the piping.

During plant operation, the reactor coolant water chemistry is monitored and maintained within very specific limits. Contaminant concentrations are kept below the thresholds known to be conducive to stress corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. For example, during normal power operation, oxygen concentration in the RCS is expected to be in the parts-per-billion (ppb) range by controlling charging flow chemistry and maintaining hydrogen in the reactor coolant at specified concentrations.

Halogen concentrations are also stringently controlled by maintaining concentrations of chlorides and fluorides within the specified limits. Thus, during plant operation, the likelihood of stress corrosion cracking is minimized.

The potential susceptibility to primary water stress corrosion cracking (PWSCC) in materials such as alloy 82/182 in the dissimilar metal welds in the Comanche Peak Units 1 and 2 RCS primary loop piping was investigated. The locations of the alloy 82/182 welds are at Reactor Pressure Vessel Inlet Nozzles (RPVINs) and Reactor Pressure Vessel Outlet Nozzles (RPVONs). The PWSCC susceptible alloy 82/182 welds are mitigated by the application of the Mechanical Stress Improvement Process (MSIP) on Units 1 RPV inlet and outlet nozzles. Unit 2 has completed a partial implementation of MSIP, where loop 4 cold leg and loop 2 hot leg nozzles (Location 12 and 1 as shown in Figure 3-2 for the respective loops) have achieved the required compressive residual stresses on the inner surface to mitigate PWSCC concerns for the alloy 82/182 welds. However, the remaining nozzle locations for Unit 2 have not yet applied the full MSIP process and thus are not yet credited for mitigation. This partial implementation of Unit 2 MSIP is addressed within this report.

2.2 WATER HAMMER Overall, there are no water hammer loading conditions required as a design input for the RCS for any associated stress analyses since it is designed and operated to preclude the voiding condition in normally filled lines. The reactor coolant system, including piping and primary components, is designed for normal, upset, emergency, and faulted condition transients. The design requirements are conservative relative to both the number of transients and their severity. Relief valve actuation and the associated hydraulic transients following valve opening are considered in the system design. Other valve and pump actuations are relatively slow transients with no significant effect on the system dynamic loads.

To ensure dynamic system stability, reactor coolant parameters are stringently controlled.

Temperature during normal operation is maintained within a narrow range; pressure is controlled by pressurizer heaters and pressurizer spray also within a narrow range for steady-state conditions. The flow characteristics of the system remain constant during a fuel cycle because the only governing parameters, namely system resistance and the reactor coolant pump characteristics, are controlled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system. Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping are such that no significant water hammer can occur.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 2.3 LOW CYCLE AND HIGH CYCLE FATIGUE An assessment of the low cycle fatigue loadings was carried out as part of this study in the form of a fatigue crack growth analysis, as discussed in Section 8.0.

High cycle fatigue loads in the system would result primarily from pump vibrations. These are minimized by restrictions placed on shaft vibrations during hot functional testing and operation.

During operation, an alarm signals the exceedance of the vibration limits. Field vibration measurements have been made on the reactor coolant loop piping in a number of plants during hot functional testing, including Comanche Peak Units 1 and 2. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for fatigue crack growth.

2.4 WALL THINNING, CREEP, AND CLEAVAGE Wall thinning by erosion and erosion-corrosion effects should not occur in the primary loop piping due to the low velocity, typically less than 1.0 ft/sec and the stainless steel material, which is highly resistant to these degradation mechanisms. The cause of wall thinning is related to high water velocity and is therefore clearly not a mechanism that would affect the primary loop piping.

Creep is typical experienced for temperatures over 700°F for stainless steel material, and the maximum operating temperature of the primary loop piping is well below this temperature value; therefore, there would be no significant mechanical creep damage in stainless steel piping.

Cleavage type failures are not a concern for the operating temperatures and the stainless steel material used in the primary loop piping.

2.5 REFERENCES

2-1 Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants, NUREG-0531, U.S. Nuclear Regulatory Commission, February 1979.

2-2 Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission, September 1980.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 PIPE GEOMETRY AND LOADING

3.1 INTRODUCTION

TO METHODOLOGY The general approach is discussed first. As an example, a segment of the primary coolant hot leg pipe is shown in Figure 3-1. The as-built outside diameter and minimum wall thickness of the pipe are 33.87 in.

and 2.33 in., respectively, as shown in the figure. The normal stresses at the weld locations are from the load combination procedure discussed in Section 3.3 whereas the faulted loads are as described in Section 3.4. The components for normal loads are pressure, deadweight, and thermal expansion (Tables 3-1 and 3-2). For Unit 1, MSIP loads are included. For Unit 2, conditions with and without MSIP applied are evaluated to account for the partial implementation as described in Section 2.1. An additional component, Safe Shutdown Earthquake (SSE), is considered for faulted loads (Tables 3-3 and 3-4). Tables 3-1 through 3-4 show the enveloping loads for Comanche Peak Units 1 and 2. As seen from Tables 3-3 and 3-4, the highest stressed location in the entire Comanche Peak Units 1 and 2 reactor coolant loops is at the reactor vessel outlet nozzle to pipe weld (Location 1) and reactor coolant pump outlet nozzle to pipe weld (Location 10).

These are two of the locations at which leak-before-break are to be established. Essentially a circumferential flaw is postulated to exist at these locations which are subjected to both the normal loads and faulted loads to assess leakage and stability, respectively. The loads (developed below) at Unit 1 location 1 are also given in Figure 3-1.

Since the primary loop piping are made of Cast Austenitic Stainless Steel (CASS) material A351-CF8A that is susceptible to thermal aging at the reactor operating temperature, the analyses also consider the associated reductions in fracture toughness. Fracture toughness properties are determined in Section 4.0. The four (two for each unit) most critical locations among the entire primary loop are identified after the full analysis is completed (see Section 5.0). At these locations, leak rate evaluations (Section 6.0) and fracture mechanics evaluations (Section 7.0) are performed per the guidance of References 3-1 and 3-2.

For global failure mechanism, all critical locations are evaluated from the impact of moment loads on the reactor coolant loop metal-base A351-CF8A stainless steel material properties which present a limiting condition due to their lower tensile properties (compared to the welding materials), as shown in Section 4.2.

For local stability mechanism, all critical locations are evaluated from the impact of thermal aging effects for the 60-year plant operating service. The CF8A material properties present a limiting condition not only due to their tensile properties in unaged condition but also the aged material fracture toughness and aged tearing modulus property, as shown in Section 4.3.

Fatigue crack growth (Section 8.0) assessment and stability margins are also evaluated (Section 9.0). All the weld locations considered for the LBB evaluation are those shown in Figure 3-2.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 3.2 CALCULATION OF LOADS AND STRESSES The stresses due to axial loads and bending moments are calculated by the following equation:

(3-1)

where,

= stress, ksi F = axial load, kips M = bending moment, in-kips A = pipe cross-sectional area, in2 Z = section modulus, in3 The total moments for the desired loading combinations are calculated by the following equation:

(3-2)

M M M M

where, M = total moment for required loading MX = X component of moment (torsion)

MY = Y component of bending moment MZ = Z component of bending moment NOTE: X-axis is along the center line of the pipe.

The axial load and bending moments for leak rate predictions and crack stability analyses are computed by the methods to be explained in Sections 3.3 and 3.4.

3.3 LOADS FOR LEAK RATE EVALUATION The normal operating loads for leak rate predictions are calculated by the following equations:

F = FDW + FTH + FMSIP + FP (3-3)

MX = (MX)DW + (MX)TH + (MX)MSIP (3-4)

MY = (MY)DW + (MY)TH + (MY)MSIP (3-5)

MZ = (MZ)DW + (MZ)TH + (MZ)MSIP (3-6)

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 The subscripts of the above equations represent the following loading cases:

DW = deadweight TH = normal thermal expansion MSIP = mechanical stress improvement process P = load due to internal pressure This method of combining loads is often referred to as the algebraic sum method (References 3-1 and 3-2).

The loads based on this method of combination are provided in Tables 3-1, 3-2A, and 3-2B at all the weld locations identified in Figure 3-2. These loads bound both loops for Comanche Peak Units 1 and 2.

3.4 LOAD COMBINATION FOR CRACK STABILITY ANALYSES In accordance with Standard Review Plan 3.6.3 (References 3-1 and 3-2), the margin in terms of applied loads needs to be demonstrated by crack stability analysis. Margin on loads of 1.4 (2) can be demonstrated if normal plus Safe Shutdown Earthquake (SSE) are applied. The 1.4 (2) margin should be reduced to 1.0 if the deadweight, thermal expansion, internal pressure, pressure expansion, Safe Shutdown Earthquake (SSE) loads are combined based on individual absolute values as shown below.

The absolute sum of loading components is used for the LBB analysis which results in higher magnitude of combined loads and thus satisfies a margin on loads of 1.0. The absolute summation of loads is shown in the following equations:

F = FDW + FTH + FMSIP + FP + FSSE (3-7)

MX = (MX)DW + (MX)TH + (MX)MSIP + (MX)SSE (3-8)

MY = (MY)DW + (MY)TH + (MY)MSIP + (MY)SSE (3-9)

MZ = (MZ)DW + (MZ)TH + (MZ)MSIP + (MZ)SSE (3-10) where subscript SSE refers to safe shutdown earthquake.

The loads so determined are used in the fracture mechanics evaluations (Section 7.0) to demonstrate the LBB margins at the locations established to be the governing locations. These loads at all the weld locations (see Figure 3-2) are given in Tables 3-3, 3-4A and 3-4B. The loads in these tables bound both loops for Comanche Peak Units 1 and 2.

3.5 REFERENCES

3-1 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

3-2 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 Table 3-1 Dimensions, Normal Loads and Stresses for Comanche Peak Unit 1 Location Outside Minimum Axial Loadb Moment Total Stress Weld Pointsa Diameter (in) Thickness (in) (lbs) (in-lbs) (psi) 1 33.87 2.33 1481520 23946856 20467 2 33.87 2.33 1481520 11779467 13328 3 37.382 3.086 1533928 20940430 12555 4 36.17 2.48 1704928 3664131 8265 5 36.17 2.48 1698688 3737998 8277 6 36.17 2.48 1693168 3480501 8132 7 36.17 2.48 1718128 2337840 7675 8 36.17 2.48 1718128 2229539 7623 9 37.5 3.145 1759098 2547169 6128 10 32.13 2.21 1356577 12873162 15379 11 32.13 2.21 1356577 9835793 13291 12 32.13 2.21 1352067 11973154 14739 Notes:

a. See Figure 3-2

b. Included Pressure

c. The impact of MSIP has been included in the calculated loads and stresses for Unit 1.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5 Table 3-2A Dimensions, Normal Loads and Stresses for Comanche Peak Unit 2 (without MSIP)

Location Outside Minimum Axial Loadb Moment Total Stress Weld Pointsa Diameter (in) Thickness (in) (lbs) (in-lbs) (psi) 1 33.87 2.33 1491160 23797092 20421 2 33.87 2.33 1490740 11535160 13225 3 37.382 3.086 1536238 20398386 12357 4 36.17 2.48 1700358 3218758 8033 5 36.17 2.48 1695478 3124369 7968 6 36.17 2.48 1689648 3227298 7996 7 36.17 2.48 1709318 1992241 7474 8 36.17 2.48 1709318 1145663 7065 9 37.5 3.145 1716598 1343529 5556 10 32.13 2.21 1351107 12348346 14992 11 32.13 2.21 1351107 9585965 13093 12 32.13 2.21 1350007 10766500 13899 Notes:

a. See Figure 3-2

b. Included Pressure Pipe Geometry and Loading January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-6 Table 3-2B Dimensions, Normal Loads and Stresses for Comanche Peak Unit 2 (with MSIP)

Location Outside Minimum Axial Loadb Moment Total Stress Weld Pointsa Diameter (in) Thickness (in) (lbs) (in-lbs) (psi) 1 33.87 2.33 1481490 23820325 20393 2 33.87 2.33 1481070 11473829 13147 3 37.382 3.086 1530408 20137247 12240 4 36.17 2.48 1706018 3564323 8221 5 36.17 2.48 1699868 3327967 8084 6 36.17 2.48 1694038 3421635 8107 7 36.17 2.48 1718178 2326742 7670 8 36.17 2.48 1718178 2231838 7624 9 37.5 3.145 1712208 2536220 5986 10 32.13 2.21 1356637 12826483 15347 11 32.13 2.21 1356637 9809574 13273 12 32.13 2.21 1351437 11062722 14110 Notes:

a. See Figure 3-2

b. Included Pressure Pipe Geometry and Loading January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-7 Table 3-3 Dimensions, Faulted Loads and Stresses for Comanche Peak Unit 1 Location Outside Minimum Axial Loadb Moment Total Stress Weld Pointsa Diameter (in) Thickness (in) (lbs) (in-lbs) (psi) 1 33.87 2.33 2182660 34644031 29780 2 33.87 2.33 2146660 28188747 25837 3 37.382 3.086 2341728 39500056 22024 4 36.17 2.48 1871728 20383744 16976 5 36.17 2.48 1836428 15530847 14498 6 36.17 2.48 1830908 15063002 14251 7 36.17 2.48 1859868 11360863 12573 8 36.17 2.48 1859868 10260184 12041 9 37.5 3.145 1847878 14786906 10934 10 32.13 2.21 1517577 19147971 20467 11 32.13 2.21 1517577 16025977 18321 12 32.13 2.21 1516067 16471917 18620 Notes:

a. See Figure 3-2

b. Included Pressure

c. The impact of MSIP has been included in the calculated loads and stresses for Unit 1.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-8 Table 3-4A Dimensions, Faulted Loads and Stresses for Comanche Peak Unit 2 (without MSIP)

Location Outside Minimum Axial Loadb Moment Total Stress Weld Pointsa Diameter (in) Thickness (in) (lbs) (in-lbs) (psi) 1 33.87 2.33 1986250 27606141 24800 2 33.87 2.33 2000250 17295652 18811 3 37.382 3.086 2193008 34291660 19601 4 36.17 2.48 1901968 26116995 19861 5 36.17 2.48 1828938 17878514 15603 6 36.17 2.48 1816268 8798479 11169 7 36.17 2.48 1917228 12989223 13578 8 36.17 2.48 1912928 12456080 13304 9 37.5 3.145 1836098 11600242 9716 10 32.13 2.21 1676327 25816471 25815 11 32.13 2.21 1683387 16531233 19466 12 32.13 2.21 1652217 20666311 22159 Notes:

a. See Figure 3-2

b. Included Pressure Pipe Geometry and Loading January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-9 Table 3-4B Dimensions, Faulted Loads and Stresses for Comanche Peak Unit 2 (with MSIP)

Location Outside Minimum Axial Loadb Moment Total Stress Weld Pointsa Diameter (in) Thickness (in) (lbs) (in-lbs) (psi) 1 33.87 2.33 1995920 27918331 25025 2 33.87 2.33 2009920 17730924 19109 3 37.382 3.086 2198838 34697834 19773 4 36.17 2.48 1907628 26732400 20180 5 36.17 2.48 1833328 18394018 15869 6 36.17 2.48 1820658 9208904 11384 7 36.17 2.48 1926088 13657247 13935 8 36.17 2.48 1921788 13228214 13711 9 37.5 3.145 1840488 13205570 10325 10 32.13 2.21 1681857 26753318 26485 11 32.13 2.21 1688917 17192341 19947 12 32.13 2.21 1653647 21489338 22731 Notes:

a. See Figure 3-2

b. Included Pressure Pipe Geometry and Loading January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-10 ODa = 33.87 in Unit 1 Location 1 ta = 2.33 in Normal Loadsa Faulted Loadsb Forcec: 1482 kips Forcec: 2183 kips Bending Moment: 23947 in-kips Bending Moment: 34644 in-kips a

See Table 3-1 b

See Table 3-3 c

Includes the force due to a pressure of 2235 psig Figure 3-1 Hot Leg Coolant Pipe Pipe Geometry and Loading January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-11 Figure 3-2 Comanche Peak Units 1 and 2 RCL Weld Locations Pipe Geometry and Loading January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4 MATERIAL CHARACTERIZATION 4.1 PRIMARY LOOP PIPE AND FITTINGS MATERIALS The Comanche Peak Units 1 and 2 primary loop pipe and elbow fitting material is A351-CF8A.

4.2 TENSILE PROPERTIES The Certified Materials Test Reports (CMTRs) specifically for the Comanche Peak Units 1 and 2 Reactor Coolant Loop Lines A351-CF8A material are used to establish tensile properties for the Leak-Before-Break analyses. Table 4-1 provides tensile properties for Unit 1 and Table 4-2 provides tensile properties for Unit 2.

For the dissimilar metal weld alloy 82/182 locations, typical mechanical properties are shown in Table 4-12.

The tensile properties are determined based on the applicable operating temperatures due to stretch power uprating (SPU) program per PCWG-06-35, Revision 1 (Reference 4-3), i.e.: 620.4F for Hot Leg, 557.6F for Crossover Leg and 558.0F for Cold Leg.

For A351-CF8A material heats, as shown in Tables 4-1 and 4-2, some CMTR data are available for the 70F and 650F test temperatures, but mostly are only available for 70F (room test temperature). The representative properties at 620.4°F, 557.6°F and 558.0°F are established from the tensile properties either at 70F or 650F given in Tables 4-1 and 4-2 by utilizing Section II of the 2007 ASME Boiler and Pressure Vessel Code (Reference 4-1) provided in Table 4-3. To obtain conservatisms in the calculation, the following steps are taken to determine the minimum Sy, average Sy, and minimum Su values as summarized in Table 4-4:

1. for each material heat, material property interpolations are based on the CMTR properties taken at 70F and 650F temperatures, to ultimately determine the minimum yield strength (Sy) value from all material heats.

2. for each material heat, material property interpolations are based on the CMTR properties taken at 70F and 650F temperatures, whichever numbers that provide greater Sy values, to ultimately determine the average yield strength (Sy) value from all material heats.

3. no CMTRs provide ultimate strength test values at 650 °F, therefore material property interpolations are based on the CMTR properties taken at 70F temperature to ultimately determine the minimum ultimate strength (Su) value from all material heats.

Code tensile properties at temperatures for the operating conditions considered in this LBB analysis are obtained by linear interpolation of tensile properties provided in the Code. Ratios of the Code tensile properties at the operating temperatures to the corresponding properties at the CMTR temperature are then applied to obtain the Comanche Peak Units 1 and 2 line-specific properties at operating temperatures.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 It should be noted that there is no significant impact by using the 2007 ASME Code Section II edition for material properties for the LBB analysis, as compared to the Comanche Peak ASME Code of record, i.e.:

ASME Code Section III 1977 Edition up to Summer 1979 Addenda.

Material modulus of elasticity is also interpolated from ASME Code values for the operating temperatures considered, and Poissons ratio is taken as 0.3. The temperature-dependent material properties from the ASME Code are shown in Table 4-3. The average and lower bound yield strengths, ultimate strengths, and elastic moduli for the pipe material at applicable operating temperatures are tabulated in Table 4-4.

4.3 FRACTURE TOUGHNESS PROPERTIES The pre-service fracture toughness (J) of cast austenitic stainless steel (CASS) that are of interest are in terms of JIc (J at Crack Initiation) and have been found to be very high at 600F. [

]a,c,e However, cast stainless steel is susceptible to thermal aging at the reactor operating temperature, that is, about 290°C (550°F). Thermal aging of cast stainless steel results in embrittlement, that is, a decrease in the ductility, impact strength, and fracture toughness of the material. Depending on the material composition, the Charpy impact energy of a cast stainless steel component could decrease to a small fraction of its original value after exposure to reactor temperatures during service.

In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials. The ANL research program measured mechanical properties of cast stainless steel materials after they had been heated in controlled ovens for long periods of time. ANL compiled a database, both from data within ANL and from international sources, of about 85 compositions of cast stainless steel exposed to a temperature range of 290-400C (550-750F) for up to 58,000 hours0 days 0 hours 0 weeks 0 months (6.5 years). In 2015 the work done by ANL was augmented, and the fracture toughness database for CASS materials was aged to 100,000 hours0 days 0 hours 0 weeks 0 months at 290-350°C (554-633°F). The methodology for estimating fracture properties has been extended to cover CASS materials with a ferrite content of up to 40%. From this database (NUREG/CR-4513, Revision 2), ANL developed correlations for estimating the extent of thermal aging of cast stainless steel (Reference 4-2).

ANL developed the fracture toughness estimation procedures by correlating data in the database conservatively. After developing the correlations, ANL validated the estimation procedures by comparing the estimated fracture toughness with the measured value for several cast stainless steel plant components removed from actual plant service. The procedure developed by ANL was used to calculate the end of life fracture toughness values for this analysis. The ANL research program was sponsored and the procedure was accepted by the NRC.

Based on Reference 4-2, the fracture toughness correlations used are the fully aged conditions provided for the A351-CF8A material.

It is noted that both Revision 1 and Revision 2 of NUREG/CR-4513 were considered in evaluating the thermal aging of the Comanche Peak CASS materials. Tables 4-10 and 4-11 provide a comparison of the JIc values for each CASS material heat calculated using both Revision 1 and Revision 2 of NUREG/CR-4513. While Revision 1 provides more JIc limiting for material heats, it was found that Revision 2 (Reference 4-2) resulted in the most limiting fracture toughness (Jmax and Tmat) values for the critical material heats identified in Table 4-9.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 Fracture Toughness Properties of Static Cast Elbows (CF8A)

The susceptibility of the material to thermal aging increases with increasing ferrite contents, and the CF8A material shows increased susceptibility to thermal aging.

The chemical compositions of the Comanche Peak Units 1 and 2 primary loop piping and elbow fitting material are available from CMTRs, and shown in Tables 4-5 and 4-6. The following equations are taken from Reference 4-2 Revision 2 and applicable for CF8A type material.

Creq = Cr + 1.21(Mo) + 0.48(Si) - 4.99 = (Chromium equivalent) [4-1]

Nieq = (Ni) + 0.11(Mn) - 0.0086(Mn)2 + 18.4(N) + 24.5(C) + 2.77 = (Nickel equivalent) [4-2]

Note: N is not included in CMTR. Value of 0.04 is assumed per Reference 4-2.

c = 100.3 (Creq / Nieq)2 -170.72(Creq / Nieq)+74.22 [4-3]

Note: CMTRs for the pipes (both Unit 1 and Unit 2) do not provide values of %Mo. A limiting value of either 0.00% or 0.50% is used, whichever results in the most limiting fracture toughness properties.

The elements are in percent weight and c is ferrite in percent volume.

The saturation room temperature (RT at 77F) impact energies of the cast stainless steel materials are determined from the chemical compositions available from CMTRs and shown in Tables 4-7 and 4-8.

For CF8A, the saturation value of RT impact energy Cvsat (J/cm2) is the lower value determined from log10Cvsat = 1.15 + 1.36 exp (-0.035) [4-4]

where the material parameter is expressed as

= c (Cr + Si)(C + 0.4N) [4-5]

and from log10Cvsat = 5.64 - 0.006c - 0.185Cr + 0.273Mo - 0.204Si+ 0.044Ni - 2.12(C + 0.4N) [4-6]

Notes: for %Mo, see a note provided for Equation 4-3. The calculated Cvsat value is conservatively used to represent Cv value in the following equations.

The J-R curve at RT, for static-cast CF8A steel (elbows) is given by Jd = 49 (Cv)0 52(a)n [4-7a]

for centrifugal-cast CF8A steel (pipes) is given by Jd = 57 (Cv)0 52(a)n [4-7b]

n = 0.18 + 0.10 log10 (Cv) [4-8]

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-4 where Jd is the deformation J in kJ/m2 and a is the crack extension in mm.

The J-R curve at 290-320C (554-608F), for static-cast CF8A steel (elbows) is given by Jd = 102 (Cv)0 28(a)n [4-9a]

for centrifugal-cast CF8A steel (pipes) is given by Jd = 134 (Cv)0 28(a)n [4-9b]

n = 0.17 + 0.09 log10 (Cv) [4-10]

where Jd is the deformation J in kJ/m2 and a is the crack extension in mm.

JIc and Jmax Calculations:

The crack extension for Jd at initiation is calculated using the ASTM E813-85 procedures, Jd at initiation (JIc) is defined on the 0.2 mm offset line (10 mm for Jmax).

Tmat Calculations:

The material shearing modulus, Tm, at aged condition is calculated as follows:

Tmat = dJ/da x E/(fa)2 Where :

E = Elastic Modulus at operating temperature, psi.

fa = aged flow stress (per Reference 4-2)

Japp and Tapp Calculations:

The critical heat for CF8A with lowest fracture toughness property and lowest tearing modulus values from Tables 4-7 and 4-8 are summarized in Table 4-9.

The applied J Integral value, Japp, is calculated and compared to the JIc and Jmax values in Table 7-1 for Units 1 and 2.

Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-5 Consideration of Dissimilar Metal Weld Material Profiles Comanche Peak Units 1 and 2 reactor coolant loop piping and elbow fittings are made of A351-CF8A material. The A351-CF8A material is susceptible to the thermal aging at the reactor operating temperatures.

Both the RPV inlet and outlet nozzles are joined to the piping with a forged stainless steel safe-end and a dissimilar metal (DM) weld of Alloy 82/182 weld material.

For local failure mechanism, the RPV inlet and outlet safe-end locations are evaluated using the cast stainless steel material properties (A351-CF8A) as shown in Table 4-9 which present a limiting condition due to the thermal aging effects. As stated in Reference 4-4, The fracture resistance of Alloy 82 and 52 welds have been investigated by conducting fracture toughness J-R curve tests at 24-338 °C in deionized water []. The results indicate that these welds exhibit high fracture toughness in air and high-temperature water (>93 °C).

Since nickel alloys are known to have high toughness properties and because the CF8A CASS base material of the RCL hot and cold legs are susceptible to thermal aging degradation of the fracture toughness, it is determined that the CF8A CASS base material presents the most limiting condition.

For the DMW locations, the evaluation is represented by J-integral for location 1 (RPV outlet nozzle safe ends) based on A351-CF8A CASS base metal property that presents the most limiting condition.

4.4 REFERENCES

4-1 ASME Boiler and Pressure Vessel Code Section II, 2007 Edition through 2008 Addenda.

4-2 O. K. Chopra, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, NUREG/CR-4513, U.S. Nuclear Regulatory Commission, Washington, DC, both Revision 1, May 1994 (ANL-93/22) and Revision 2, May 2016 (ANL-15/08), plus errata for Revision 2, March 2021.

4-3 Westinghouse Letter PCWG-06-35, Revision 1,

Subject:

Comanche Peak Units 1 & 2 (TBX/TCX):

Approval of Category III (for Contract) PCWG Parameters to Support the Uprate Program, October 3, 2006.

4-4 NUREG/CR-6721, Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds", April 2001 Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-6 Notes: L = Loop; H = Hot Leg; X = Crossover Leg; C = Cold Leg; EL = Elbow; N/A= Not applicable Table 4-1 Measured Tensile Properties for Comanche Peak Unit 1 Primary Loop Pipes and Elbows (A351-CF8A)

At Room Temperature At 650°F Heat/Serial RCL No. Yield Ultimate Yield Ultimate Strength (psi) Strength (psi) Strength (psi) Strength (psi)

L1-H 153884 39500 79150 N/A N/A L2-H 153885 40500 81750 N/A N/A L3-H 153882 39650 82650 N/A N/A L4-H 153883 42700 81300 N/A N/A L1-H-EL 3-3249/1620 36625 80447 22615 N/A L2-H-EL 3-3271/1724 36511 84145 24749 N/A L3-H-EL 3-3276/1931 35345 83448 24037 N/A L4-H-EL 3-3278/1932 39726 84032 24464 N/A L1-X 153701 Pc-1 44450 85850 N/A N/A L1-X 153700 Pc-2 42000 83100 N/A N/A L2-X 153701 Pc-2 44450 85850 N/A N/A L2-X 153701 Pc-3 44450 85850 N/A N/A L3-X 153701 Pc-4 44450 85850 N/A N/A L3-X 153700 Pc-3 42000 83100 N/A N/A L4-X 153700 Pc-4 42000 83100 N/A N/A L4-X 153894 Pc-4 38300 85400 N/A N/A L1-X-EL 3-3319/2665 37208 81827 24037 N/A L1-X-EL 3-3378/0211 39399 85425 21762 N/A L1-X-EL40 3-3325 43808 89835 26171 N/A L2-X-EL 3-3572/2799 40821 85482 N/A N/A L2-X-EL 3-3380/0384 43922 92494 N/A N/A L2-X-EL40 3-3548 35231 82524 N/A N/A L3-X-EL 3-3357/0638 39754 89365 27451 N/A L3-X-EL 3-3576/3229 39171 86350 N/A N/A L3-X-EL40 3-3405 40252 89251 N/A N/A L4-X-EL 3-3619/4510 45614 96150 N/A N/A L4-X-EL 3-3627/4509 42770 88753 N/A N/A Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-7 Table 4-1 Measured Tensile Properties for Comanche Peak Unit 1 Primary Loop Pipes and Elbows (A351-CF8A)

At Room Temperature At 650°F Heat/Serial RCL No. Yield Ultimate Yield Ultimate Strength (psi) Strength (psi) Strength (psi) Strength (psi)

L4-X-EL40 3-3381 41034 85198 N/A N/A L1-C 154881 45600 82400 N/A N/A L2-C 153891 44650 85800 N/A N/A L3-C 153887 40800 81100 N/A N/A L4-C 153889 41400 78000 N/A N/A L1-C-EL 3-3249/1600 37976 80362 25744 N/A L2-C-EL 3-3249/1620 38972 84032 22615 N/A L3-C-EL 3-3271/1724 36511 84145 24749 N/A L4-C-EL 3-3276/1901 38247 80077 21762 N/A Table 4-2 Measured Tensile Properties for Comanche Peak Unit 2 RCL Pipes and Elbows (A351-CF8A)

At Room Temperature At 650°F Heat/Serial RCL No. Yield Ultimate Yield Ultimate Strength (psi) Strength (psi) Strength (psi) Strength (psi)

L1-H 156517 Pc-1 40800 82500 N/A N/A L2-H 156570 43950 83400 N/A N/A L3-H 156571 42900 83750 N/A N/A L4-H 156572 46600 83300 N/A N/A L1-H-EL 3-3733/1834 43580 88441 N/A N/A L2-H-EL 3-3719/1833 44334 90744 N/A N/A L3-H-EL 3-4025/1831 40394 86905 N/A N/A L4-H-EL 3-3703/1576 41390 88441 N/A N/A L1-X 156374 Pc-1 38600 82400 N/A N/A L1-X 156374 Pc-1 38600 82400 N/A N/A L2-X 156374 Pc-2 38600 82400 N/A N/A L2-X 156374 Pc-2 38600 82400 N/A N/A L3-X 156375 Pc-1 41600 84150 N/A N/A L3-X 156375 Pc-1 41600 84150 N/A N/A Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-8 Table 4-2 Measured Tensile Properties for Comanche Peak Unit 2 RCL Pipes and Elbows (A351-CF8A)

At Room Temperature At 650°F Heat/Serial RCL No. Yield Ultimate Yield Ultimate Strength (psi) Strength (psi) Strength (psi) Strength (psi)

L4-X 156375 Pc-2 41600 84150 N/A N/A L4-X 156375 Pc-2 41600 84150 N/A N/A L1-X-EL 3-4047/3172 38830 86336 N/A N/A L1-X-EL 3-3704/1007 44732 92608 N/A N/A L-X-EL40* 3-3506 38588 88213 N/A N/A L2-X-EL 3-4020/5762 37550 82353 N/A N/A L2-X-EL 3-3452-1166 43225 91911 N/A N/A L-X-EL40* 3-3489 39171 88441 N/A N/A L3-X-EL 3-3950/1828 45230 91598 N/A N/A L3-X-EL 3-3923/1829 42101 90460 N/A N/A L-X-EL40* 3-3594 37777 89138 N/A N/A L4-X-EL 3-3588/6011 38944 88327 N/A N/A L4-X-EL 3-3729/1939 45031 90987 N/A N/A L-X-EL40* 3-3612 41959 88469 N/A N/A L1-C 156362 40100 83150 N/A N/A L2-C 156363 45850 85300 N/A N/A L3-C 156364 43050 84250 N/A N/A L4-C 156574 45700 84050 N/A N/A L1-C-EL 3-3657/5292 45202 90531 N/A N/A L2-C-EL 3-3719/1832 43523 89835 N/A N/A L3-C-EL 3-3695/2342 43225 88782 N/A N/A L4-C-EL 3-3718/1832 41817 86706 N/A N/A Note: L-X-EL40* = Cross-Over Leg Elbows (40 deg.) provided in the table are not specifically associated with any loop number in Unit 2. For example, heat number 3-3506 is not necessarily located in loop 1 because it is listed under L1-X-EL, but 3-3506 could be located either in loop 1, 2, 3, or 4. However, it does not matter, to obtain minimum and average and tensile properties for CF8A from Units 1 and 2 in Table 4-4, in which loop the heat 3-3506 is located.

Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-10 Table 4-4 Tensile Properties for Comanche Peak Units 1 and 2 A351-CF8A Materials at Operating Temperatures T** Avg. Min. Sy Min. Su E Poisson Materials*

(F) Sy (psi) (psi) (psi) (psi) ratio 620.4 25330 21506 66918 25198000 0.3 A351-CF8A 557.6 25950 22111 69181 25554400 0.3 558.0 26758 22759 65945 25552000 0.3

Material tensile properties are in unaged condition.

A351-CF8A tensile properties (Sy, and Su) increase due to thermal aging.

- The upper bound temperatures are used in this calculation note to reflect the operating temperature due to SPU program. Hot Leg Operating Temperature = 620.4°F, Cross-over Leg Operating Temperature = 557.6°F, Cold Leg Operating Temperature =558.0°F.

Material Characterization January 2023 WCAP-10528-NP Revision 3

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a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-11 Table 4-5 Comanche Peak Unit 1 CF8A Chemical Composition Material Characterization January 2023 WCAP-10528-NP Revision 3

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a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-12 Table 4-6 Comanche Peak Unit 2 CF8A Chemical Composition Material Characterization January 2023 WCAP-10528-NP Revision 3

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a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-13 Table 4-7 Comanche Peak Unit 1 CF8A Fracture Toughness and Tearing Modulus Properties Material Characterization January 2023 WCAP-10528-NP Revision 3

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a,c,e WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-14 Table 4-8 Comanche Peak Unit 2 CF8A Fracture Toughness and Tearing Modulus Properties Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-15 a,c,e Table 4-9 Comanche Peak Units 1 and 2 CF8A with Lowest Fracture Toughness Properties Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-16 a,c,e Table 4-10 JIC Limiting Values NUREG/CR-4513 either per Rev. 1 or per Rev. 2 for Unit 1 Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-17 a,c,e Table 4-11 JIC Limiting Values NUREG/CR-4513 either per Rev. 1 or per Rev. 2 for Unit 2 Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-18 Table 4-12 Alloy 82/182 Tensile Properties Ultimate Typical Yield* Modulus of Temperature (°F) Strength, Su Strength, Sy (ksi) Elasticity, E, (ksi)

(ksi) 600 49.700 84.000 28700 620.4 49.659 83.714 28618 650 49.600 83.300 28500 Note:

Typical Yield Strength is considered as the average yield strength Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-19 Figure 4-1 Pre-Service J vs. a for Cast Stainless Steel at 600F Material Characterization January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5 CRITICAL LOCATION AND EVALUATION CRITERIA 5.1 CRITICAL LOCATIONS The governing or critical locations for the LBB evaluation are established based on the fracture toughness properties of the metal-base at the weld points and also on the basis of pipe geometry, welding process, operating temperature, operating pressure, and the highest faulted stresses at the welds.

The RCL weld points applicable for Comanche Peak Unit 1 and 2 are shown in Figure 3-2.

Critical locations for LBB Evaluation are determined based on the maximum faulted stresses for Unit 1 (weld point 1) and Unit 2 (weld point 10). Both weld points 1 and 10 are evaluated for Units 1 and 2.

For LBB evaluation, Table 5-1 shows the critical locations bounding both Comanche Peak Units 1 and 2.

Figure 5-1 shows the locations of the critical welds for Comanche Peak Units 1 and 2.

5.2 EVALUATION CRITERIA As will be discussed later, fracture mechanics analyses are made based on local failure mechanism as described in Section 7.1 and based on global failure mechanism as described in Section 7.2.

For local failure mechanism, stability analysis is performed using J-integral evaluation method with the criteria as follows:

(1) If Japp < JIc, then the crack will not initiate, and the crack is stable; (2) If Japp > JIc; and Tapp < Tmat and Japp < Jmax, then the crack is stable.

Where:

Japp = Applied J JIc = J at Crack Initiation Tapp = Applied Tearing Modulus Tmat = Material Tearing Modulus Jmax = Maximum J value of the material For global failure mechanism, the stability analysis is performed using limit load method based on loads and postulated flaw sizes related to leakage, with the criteria as follows:

Margin of 10 on the Leak Rate Margin of 2.0 on Flaw Size Margin of 1.0 on Loads (using the absolute summation method for faulted load combination).

Critical Location and Evaluation Criteria January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-2 Table 5-1 Critical Analysis Locations for Comanche Peak Units 1 and 2 RCL Lines Operating Operating Maximum Do Pipe Thickness Welding Weld Pts. Pressure Temperature Faulted Stress (in) (in) Process (psia) (°F) (psi) 1 33.87 2.33 SAW 2250 620.4 29780 (Unit 1) 10 32.13 2.21 SAW 2250 558.0 20467 (Unit 1) 1 33.87 2.33 SAW 2250 620.4 24800 (Unit 2 without MSIP) 10 32.13 2.21 SAW 2250 558.0 25815 (Unit 2 without MSIP) 1 33.87 2.33 SAW 2250 620.4 25025 (Unit 2 with MSIP) 10 32.13 2.21 SAW 2250 558.0 26485 (Unit 2 with MSIP)

Note: Unit 1 stresses have included MSIP.

Critical Location and Evaluation Criteria January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 Figure 5-1 Schematic Diagram of Comanche Peak Units 1 and 2 Primary Loop Showing Critical Weld Locations Critical Location and Evaluation Criteria January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6 LEAK RATE PREDICTIONS

6.1 INTRODUCTION

The purpose of this section is to discuss the method which is used to predict the flow through postulated through-wall cracks and present the leak rate calculation results for through-wall circumferential cracks.

6.2 GENERAL CONSIDERATIONS The flow of hot pressurized water through an opening to a lower back pressure causes flashing which can result in choking. For long channels where the ratio of the channel length, L, to hydraulic diameter, DH, (L/DH) is greater than [

]a,c,e 6.3 CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by [

]a,c,e The flow rate through a crack was calculated in the following manner. Figure 6-1 from Reference 6-2 was used to estimate the critical pressure, Pc, for the primary loop enthalpy condition and an assumed flow. Once Pc was found for a given mass flow, the [ ]a,c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [

]a,c,e Therefore, this method will yield the two-phase pressure drop due to momentum effects as illustrated in Figure 6-3, where Po is the operating pressure. Now using the assumed flow rate, G, the frictional pressure drop can be calculated using (6-1) where the friction factor f is determined using the [ ]a,c,e The crack relative roughness, ,

was obtained from fatigue crack data on stainless steel samples. The relative roughness value used in these calculations was [ ]a,c,e The frictional pressure drop using equation 6-1 is then calculated for the assumed flow rate and added to the [ ]a,c,e to obtain the total pressure drop from the primary system to the atmosphere. That is, for the primary loop:

Absolute Pressure - 14.7 = [ ]a,c,e (6-2) for a given assumed flow rate G. If the right-hand side of equation 6-2 does not agree with the pressure difference between the primary loop and the atmosphere, then the procedure is repeated until equation 6-2 is satisfied to within an acceptable tolerance which in turn leads to flow rate value for a given crack size.

Leak Rate Predictions January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 6.4 LEAK RATE CALCULATIONS Leak rate calculations were made as a function of crack length at the governing locations previously identified in Section 5.1. The normal operating loads of Tables 3-1 and 3-2 were applied in these calculations. The crack opening areas were estimated using the method of Reference 6-3, and the leak rates were calculated using the two-phase flow formulation described in the preceding section. The average material properties of Section 4.0 (see Table 4-4) were used for these calculations.

The flaw sizes to yield a leak rate of 10 gpm were calculated at the governing locations and are given in Table 6-1 for Comanche Peak Units 1 and 2 CASS material and Table 6-2 for Units 1 and 2 alloy 82/182 weld material. Based on the PWSCC morphology, a conservative factor of 1.69 between the PWSCC and fatigue crack morphologies (Reference 6-4) is applied to the leakage flaw sizes for the alloy 82/182 material in Table 6-2. The flaw sizes so determined are called leakage flaw sizes (crack lengths).

Per WCAP-10527, Chapter 5 (Reference 1-7), the Comanche Peak Units 1 and 2 RCS pressure boundary leak detection system capability is 1 gpm in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Thus, to satisfy the margin of 10 on the leak rate, the flaw sizes (leakage flaw sizes) (crack lengths) are determined which yield a leak rate of 10 gpm.

6.5 REFERENCES

6-1 [

]a,c,e 6-2 M. M, El-Wakil, Nuclear Heat Transport, International Textbook Company, New York, N.Y, 1971.

6-3 Tada, H., The Effects of Shell Corrections on Stress Intensity Factors and the Crack Opening Area of Circumferential and a Longitudinal Through-Crack in a Pipe, Section II-1, NUREG/CR-3464, September 1983.

6-4 D. Rudland, R. Wolterman, G. Wilkowski, R. Tregoning, Impact of PWSCC and Current Leak Detection on Leak-Before-Break, proceedings of Conference on Vessel Head Penetration, Inspection, Cracking, and Repairs, Sponsored by the USNRC, Marriot Washingtonian Center, Gaithersburg, MD, September 29 to October 2, 2003.

Leak Rate Predictions January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-1 Flaw Sizes for Comanche Peak Units 1 and 2 Yielding a Leak Rate of 10 gpm for the RCL Lines with A351-CF8A Material Weld Points Leakage Flaw Size (in) 1 (Unit 1) 3.55 10 (Unit 1) 4.50 1

3.56 (Unit 2 without MSIP) 10 4.59 (Unit 2 without MSIP) 1 3.57 (Unit 2 with MSIP) 10 4.51 (Unit 2 with MSIP)

Notes: The flaw size in the Table 6-1 refers to the flaw length of through-wall circumferential crack. Unit 1 results have included MSIP.

Leak Rate Predictions January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 Table 6-2 Flaw Sizes for Comanche Peak Units 1 and 2 Yielding a Leak Rate of 10 gpm for the Critical Analysis Location with Alloy 82/182 Welds Leakage Flaw Size Leakage Flaw Size with 1.69 Weld Points (in) PWSCC factor applied 1

(Unit 1) 4.77 8.06 1

4.78 8.08 (Unit 2 without MSIP) 1 4.79 8.10 (Unit 2 with MSIP)

Leak Rate Predictions January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 a,c,e Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-7 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7 FRACTURE MECHANICS EVALUATION 7.1 LOCAL FAILURE MECHANISM The local mechanism of failure is primarily dominated by the crack tip behavior in terms of crack-tip blunting, initiation, extension, and final crack instability. The local stability will be assumed if the crack does not initiate at all. It has been accepted that the initiation toughness measured in terms of JIc from a J-integral resistance curve is a material parameter defining the crack initiation. If, for a given load, the calculated J-integral value is shown to be less than the JIc of the material, then the crack will not initiate. If the initiation criterion is not met, one can calculate the tearing modulus as defined by the following relation:

dJ E (7-1)

T x da where:

Tapp = applied tearing modulus E = modulus of elasticity f = 0.5 (y + u) = flow stress a = crack length y, u = yield and ultimate strength of the material, respectively Stability is said to exist when ductile tearing does not occur if Tapp is less than Tmat, the experimentally determined tearing modulus. Since a constant Tmat is assumed a further restriction is placed in Japp. Japp must be less than Jmax where Jmax is the maximum value of J for which the experimental Tmat is greater than or equal to the Tapp used.

As discussed in Section 5.2 the local crack stability criteria is a two-step process:

(1) If Japp < JIc, then the crack will not initiate, and the crack is stable; (2) If Japp > JIc; and Tapp < Tmat and Japp < Jmax, then the crack is stable.

The calculations of Japp and Tapp values for the critical locations are performed following the methodology developed in References 7-2 and 7-3. The stability results based on elastic-plastic J-integral evaluations for Comanche Peak Units 1 and 2 are provided in Table 7-1.

Fracture Mechanics Evaluation January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 7.2 GLOBAL FAILURE MECHANISM Determination of the conditions which lead to failure in stainless steel should be done with plastic fracture methodology because of the large amount of deformation accompanying fracture. One method for predicting the failure of ductile material is the plastic instability method, based on traditional plastic limit load concepts, but accounting for strain hardening and taking into account the presence of a flaw. The flawed pipe is predicted to fail when the remaining net section reaches a stress level at which a plastic hinge is formed. The stress level at which this occurs is termed as the flow stress. The flow stress is generally taken as the average of the yield and ultimate tensile strength of the material at the temperature of interest.

This methodology has been shown to be applicable to ductile piping through a large number of experiments and will be used here to predict the critical flaw size in the primary coolant piping. The failure criterion has been obtained by requiring equilibrium of the section containing the flaw (Figure 7-1) when loads are applied. The detailed development is provided in Appendix A for a through-wall circumferential flaw in a pipe with internal pressure, axial force, and imposed bending moments. The limit moment for such a pipe is given by:

]a,c,e The analytical model described above accurately accounts for the piping internal pressure as well as imposed axial force as they affect the limit moment. Good agreement was found between the analytical predictions and the experimental results (Reference 7-1). For application of the limit load methodology, the material, including consideration of the configuration, must have a sufficient ductility and ductile tearing resistance to sustain the limit load.

A stability analysis based on limit load was performed for all the critical locations (locations 1 and 10 for Units 1 and 2). For the RCL Lines of Comanche Peak Units 1 and 2, the SAW weld processes are conservatively assumed to be used for all CF8A CASS materials. The Z factor correction for SAW was applied per References 7-4 and 7-5:

Z = 1.30 [1.0 + 0.01 (OD-4)] for SAW (7-4) where OD is the outer diameter of the pipe in inches.

Fracture Mechanics Evaluation January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 The Z-factors were calculated for the critical locations, using the dimensions given in Tables 3-1 through 3-4. The applied loads were increased by the Z factors. Table 7-2 summarizes the results of the stability analyses based on limit load for the CASS locations. The leakage flaw sizes are also presented on the same table.

Comanche Peak Units 1 and 2 reactor coolant system primary loop piping contains alloy 82/182 dissimilar metal welds which are susceptible to PWSCC. Locations of the alloy 82/182 welds are at the RPV inlet and outlet nozzles (locations 12 and 1 respectively). However, critical analysis locations are determined to be location 1 for both Units 1 and 2, as shown in Figure 5-1. [

]a,c,eTable 7-3 summarizes the results of the stability analyses based on limit load for the alloy 82/182 location 1. The leakage flaw sizes are also presented on the same table.

7.3 REFERENCES

7-1 Kanninen, M. F., et. Al., Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks, EPRI NP-192, September 1976.

7-2 [

]a,c,e 7-3 [

] a,c,e 7-4 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

7-5 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

7-6 ASME Pressure Vessel and piping Division Conference Paper PVP2008-61840, Technical Basis for Revision to Section XI Appendix C for Alloy 600/82/182/132 Flaw Evaluation in Both PWR and BWR Environments, July 28-31, Chicago IL, USA.

Fracture Mechanics Evaluation January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4 Table 7-1 Stability Results for Comanche Peak Units 1 and 2 Based on J-Integral Evaluations and CASS Thermal Aging Material Calculated(1) Fracture Criteria / Limit Location Japp JIc Jmax Tapp (in-lb/in2) (in-lb/in2) (in-lb/in2) Tmat(2) a,c,e Fracture Mechanics Evaluation January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5 Table 7-2 Flaw Stability Results for Comanche Peak Units 1 and 2 RCL Lines Based on Limit Load with A351-CF8A Material Weld Points Leakage Flaw Size (in) Critical Flaw Size (in) 1 3.55 10.84 (Unit 1) 10 4.50 21.95 (Unit 1) 1 3.56 16.82 (Unit 2 without MSIP) 10 4.59 15.66 (Unit 2 without MSIP) 1 3.57 16.54 (Unit 2 with MSIP) 10 4.51 14.93 (Unit 2 with MSIP)

Note: Unit 1 results have included the impact of MSIP.

Fracture Mechanics Evaluation January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-6 Table 7-3 Flaw Stability Results for Comanche Peak Units 1 and 2 RCL Lines Based on Limit Load with Alloy 82/182 welds Weld Points Leakage Flaw Size (in) Critical Flaw Size (in) 1 8.06 32.93 (Unit 1) 1 8.08 36.84 (Unit 2 without MSIP) 1 8.10 36.66 (Unit 2 with MSIP)

Fracture Mechanics Evaluation January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-7 Figure 7-1 [ ]a,c,e Stress Distribution Fracture Mechanics Evaluation January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8 FATIGUE CRACK GROWTH ANALYSIS To determine the sensitivity of the primary coolant system to the presence of small cracks, a fatigue crack growth analysis was carried out for the [ ]a,c,e region of a typical system (see a,c,e Location [ ] of Figure 3-2). This region was selected because crack growth calculated here will be typical of that in the entire primary loop. Crack growths calculated at other locations can be expected to show minimal variation.

A[ ]a,c,e of a plant typical in geometry and operational characteristics to any Westinghouse PWR System. [

]a,c,e The normal, upset, and test conditions were considered. Circumferentially oriented surface flaws are postulated in the region, assuming the flaw was located in three different locations, as shown in Figure 8-

1. Specifically, these are:

a,c,e Fatigue crack growth rate laws were used [

]a,c,e The law for stainless steel was derived from Reference 8-1, with a very conservative correction for the R ratio, which is the ratio of minimum to maximum stress during a transient.

For stainless steel, the fatigue crack growth formula is:

Where:

da/dn = crack growth rate (inches/cycles)

Keff = Kmax(1.0 - R)0 5 R = Kmin/Kmax

[

Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-2

]a,c,e The calculated fatigue crack growth for semi-elliptic surface flaws of circumferential orientation and various depths is summarized in Table 8-2, and shows that the crack growth is very small, regardless of which material is assumed.

The reactor vessel transients and projected 60-year cycles for Comanche Peak Units 1 and 2 are shown in Table 8-3. By comparing the transients and cycles for the Comanche Peak plant specific transients and cycles presented in Table 8-3, and the generic analysis shown in Table 8-1, it is shown that there are no major differences in calculations based on the enveloping data for both 60 years of Comanche Peak plant service and the representative design data for 60 years of plant service. The representative design data is also applicable for Comanche Peak 60 year plant service. Also, if there are slight changes due to uprate transients and changes in the cycles for the 60 year design transients, they will not have a significant impact on the fatigue crack growth conclusions, since there is insignificant growth of small surface flaws as shown in Table 8-2.

The fatigue crack growth analysis is not a requirement for the LBB analysis (see References 8-4 and 8-5) since the LBB analysis is based on the postulation of through-wall flaws, whereas the FCG analysis is performed based on a surface flaw. In addition, Reference 8-6 has indicated that, the Commission deleted the fatigue crack growth analysis in the proposed rule. This requirement was found to be unnecessary because it was bounded by the crack stability analysis. This evaluation of FCG is presented as a defense-in-depth justification in support of the demonstration of LBB.

It is therefore, concluded that the fatigue crack growth analysis results shown in Table 8-2 are representative of the Comanche Peak plants fatigue crack growth for 60 Years.

Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3

8.1 REFERENCES

8-1 Bamford, W. H., Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment, Trans. ASME Journal of Pressure Vessel Technology, Vol. 101, Feb. 1979.

8-2 James, L.A., Fatigue Crack Propagation Behavior of Inconel 600, in Hanford Engineering Development Labs Report HEDL-TME-76-43, May 1976.

8-3 Hale, D. A., et. Al., Fatigue Crack Growth in Piping and RPV Steels in Simulated BWR Water Environment, Report GEAP 24098/NUREG CR-0390, Jan. 1978.

8-4 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday August 28, 1987/Notices, pp. 32626-32633.

8-5 NUREG-0800, Revision 1, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures March 2007.

8-6 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

8-7 [

]a,c,e 8-8 Westinghouse Letter, LTR-SDA-II-21-28, Revision 2, Comanche Peak Nuclear Power Plant Units 1 and 2 License Renewal - 60 Year Transient Projections for Class 1 RCS and Auxiliary Systems, Class 2 Heat Exchangers, and Class 1, 2, and 3 Valves, August 5, 2022.

Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-4 Table 8-1 Summary of Transients (Representative 60-Year Design)

Number Typical Transient Identification Number of Cycles Normal Conditions Heatup and Cooldown at 100°F/hr.

1 (pressurizer cooldown 200°F/hr.) 200*

Load Follow Cycles 2 (Unit loading and unloading at 5% of full power/min.) 18300 3 Step Load Increase and Decrease 2000 4 Large Step Load Decrease, with Steam Dump 200 5 Steady State Fluctuations 1000000 Upset Conditions 6 Loss of Load, without Immediate Turbine or Reactor Trip 80 Loss of Power 7 (blackout with natural circulation in the Reactor Coolant System) 40 Loss of Flow 8 (partial loss of flow, one pump only) 80 9 Reactor Trip from Full Power 400*

Test Conditions 10 Turbine Roll Test 10 11 Hydrostatic Test Conditions Primary Side 5 Primary Side Leak Test 50 12 Cold Hydrostatic Test 10 Note:

Number of cycles include cycles due to forced outages, plant trips, runbacks, etc.

Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 Table 8-2 Typical Fatigue Crack Growth at [ ] a,c,e (60 Years)

FINAL FLAW DEPTH (in.)

Initial Flaw Depth (in.) [ ]a,c,e [ ]a,c,e [ ]a,c,e 0.292 0.31097 0.30107 0.30698 0.300 0.31949 0.30953 0.31626 0.375 0.39940 0.38948 0.40763 0.425 0.45271 0.44350 0.47421 Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-6 Table 8-3 Summary of Transients for Comanche Peak Units 1 and 2 Number of Projected Number Transient Name Occurrences Cycles for (40-year Basis)1 60-year License2 Heatup and cooldown at 100°F/hr (pressurizer 1 200 (each) 86 cooldown 200°F/hr)

Unit loading and unloading at 5% of full power/min 2 13200 (each) 303 (RCS Loop Transients)

Step load increase and decrease of 10% of full 3 2000 (each) 2 (each) power 4 Large Step Decrease with steam dump 200 10 5 Steady state fluctuations

a. Initial fluctuations 1.5 x 105 N/A

b. Random fluctuations 3.0 x 106 N/A 6 Feedwater cycling 4000 8 7 Loop out of service

a. Normal loop shutdown 80 N/A

b. Normal loop startup 70 N/A Unit loading and unloading between 0 to 15% of 8 500 (each) 2 (each) full power 9 Boron concentration equalization 26400 552 10 Refueling (RCS Loop Transients) 80 41 11 Loss of load, without immediate reactor trip (RCS) 80 11 Loss of power (blackout with natural circulation in 12 40 12 the Reactor Coolant System) 13 Partial loss of flow (loss of one pump) 80 12 14 Reactor trip from full power (RCS Loop Transients)

a. Without cooldown 230 73

b. With cooldown, without safety injection 160 26

c. With cooldown and safety injection 10 5 15 Inadvertent reactor coolant depressurization (RCS) 20 3 16 Inadvertent startup of an inactive loop (RCS) 10 7 Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-7 Table 8-3 Summary of Transients for Comanche Peak Units 1 and 2 Number of Projected Number Transient Name Occurrences Cycles for (40-year Basis)1 60-year License2 17 Control rod drop (RCS Loop Transients) 80 5 Inadvertent Emergency Core Cooling System 18 60 10 actuation Operating Basis Earthquake (20 earthquakes of 10 19 200 2 cycles each) 20 Excessive Feedwater Flow 30 4 21 Test Conditions:

Turbine roll test 20 0 Primary side hydrostatic test 10 10 Secondary side hydrostatic test 10 10 Primary side leak test 200 200 Secondary side leak test 80 80 Tube leakage test 800 410 Notes:

1) Transients in Table 8-3 and 40-year design basis cycles are based on Comanche Peak FSAR, Chapter 3, Table 3.9N-1.

2) 60-year transient cycle projections are provided in Reference 8-8.

Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-8 Figure 8-1 Typical Cross-Section of [ ]a,c,e Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-9 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-10 Figure 8-3 Reference Fatigue Crack Growth Law for [ ]a,c,e in a Water Environment at 600°F Fatigue Crack Growth Analysis January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 9 ASSESSMENT OF MARGINS The results of the leak rates of Section 6.4 and the corresponding stability and fracture toughness evaluations of Sections 7.1 and 7.2 are used in performing the assessment of margins. Margins are shown in Table 9-1 for Units 1 and 2. All of the LBB recommended margins are satisfied. The LBB analyses results are acceptable for the License Renewal program (60 Years).

In summary, at all the critical locations relative to:

1. Flaw Size - Using faulted loads obtained by the absolute sum method, a margin of 2 or more exists between the critical flaw and the flaw having a leak rate of 10 gpm (the leakage flaw).

2. Leak Rate - A margin of 10 exists between the calculated leak rate from the leakage flaw and the plant leak detection capability of 1 gpm.

3. Loads - At the critical locations the leakage flaw was shown to be stable using the faulted loads obtained by the absolute sum method (i.e., a flaw twice the leakage flaw size is shown to be stable; hence the leakage flaw size is stable). A margin of 1 on loads using the absolute summation of faulted load combinations is satisfied.

Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Comanche Peak Units 1 and 2 with A351-CF8A CASS Material Location Leakage Flaw Size (in) Critical Flaw Size (in) Margin 1 3.55 10.84a 3.05a (Unit 1)c 3.55 7.10b >2.0b 10 4.50 21.95a 4.88a (Unit 1)c 4.50 9.00b >2.0b 1 3.56 16.82a 4.73a (Unit 2 without MSIP) 3.56 7.12b >2.0b 10 4.59 15.66a 3.41a (Unit 2 without MSIP) 4.59 9.18b >2.0b 1 3.57 16.54 4.64a (Unit 2 with MSIP) 3.57 7.14 >2.0b 10 4.51 14.93 3.31a (Unit 2 with MSIP) 4.51 9.02 >2.0b a

based on limit load b

based on J integral evaluation c

Unit 1 results have included the impact of MSIP.

Assessment of Margins January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-2 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Comanche Peak Units 1 and 2 with Alloy 82/182 Welds Location Leakage Flaw Size (in)* Critical Flaw Size (in) Margin 1

8.06 32.93 4.09 (Unit 1) 1 8.08 36.84 4.55 (Unit 2 without MSIP) 1 4.10 36.66 4.53 (Unit 2 with MSIP)

Note: *Based on a conservative factor of 1.69 PWSCC morphology Assessment of Margins January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1 10 CONCLUSIONS This report justifies the elimination of RCS primary loop pipe breaks from the structural design basis for the 60-year plant life of Comanche Peak Units 1 and 2 as follows:

a. Stress corrosion cracking is precluded by use of fracture resistant materials in the piping system and controls on reactor coolant chemistry, temperature, pressure, and flow during normal operation. Alloy 82/182 welds are present at the Comanche Peak Units 1 and 2 RPVINs and RPVONs. The alloy 82/182 welds are susceptible to PWSCC (Primary Water Stress Corrosion Cracking) and have been conservatively evaluated to consider the effects of PWSCC.

b. To mitigate PWSCC due to the existence of alloy 82/182, MSIP are applied to the Comanche Peak Units 1 RPVINs and RPVONs. Unit 2s partial implementation of MSIP has been considered within this evaluation.

c. As stated in Section 3.0, for local and global failure mechanisms, all locations are evaluated using the cast austenitic stainless steel material properties (A351-CF8A).

d. Evaluation of the RCS piping considering the thermal aging effects for the 60-year plant life period of the LR program and also the use of the most limiting fracture toughness properties ensures that each materials profile is appropriately bounded by the LBB results presented in this report.

e. Water hammer should not occur nor is it a design bases input load in the RCS piping because of system design, testing, and operational considerations.

f. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.

g. Ample margin exists between the leak rate of small stable flaws and the capability of the Comanche Peak Units 1 and 2 reactor coolant system pressure boundary Leakage Detection System.

h. Ample margin exists between the small stable flaw sizes of item (g) and larger stable flaws.

i. Ample margin exists in the material properties used to demonstrate end-of-service life (fully aged) stability of the critical flaws.

For the critical locations, flaws are identified that will be stable because of the ample margins described in g, h, and i above.

Based on the above, the Leak-Before-Break conditions and margins are satisfied for the Comanche Peak Units 1 and 2 primary loop piping. All the recommended margins are satisfied. It is therefore concluded that dynamic effects of RCS primary loop pipe breaks need not be considered in the structural design basis for Comanche Peak Units 1 and 2 Nuclear Power Plants for the 60-year plant life (License Renewal program).

Conclusions January 2023 WCAP-10528-NP Revision 3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A LIMIT MOMENT

[

]a,c,e Conclusions January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 Figure A-1 Pipe with a Through-Wall Crack in Bending Conclusions January 2023 WCAP-10528-NP Revision 3

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

WCAP-10528-NP Revision 3 Non-Proprietary Class 3

- This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Author Approval Weldon Kevin Jan-10-2023 16:29:55 Reviewer Approval Wiratmo Momo Jan-10-2023 16:44:44 Manager Approval Patterson Lynn Jan-10-2023 16:54:33 Files approved on Jan-10-2023

- - This record was final approved on 1/10/2023, 4:54:33 PM. (This statement was added by the PRIME system upon its validation)

Enclosure 4:

Affidavit CAW-23-021 3 pages follow

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-021 Page 1 of 3 Commonwealth of Pennsylvania:

County of Butler:

(1) I, Camille Zozula, Manager/Interim Director, Management Systems & Regulatory Compliance, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of WCAP-10527-P Revision 3 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-021 Page 2 of 3 (5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-021 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 6/1/2023 _____________________________

Signed electronically by Camille Zozula

CP-202300285, Response to Request for Additional Information Regarding Review of the License Renewal Application - Set 1

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

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1.1 BACKGROUND

1.3 REFERENCES

2.5 REFERENCES

3.1 INTRODUCTION

3.5 REFERENCES

4.4 REFERENCES

6.1 INTRODUCTION

6.5 REFERENCES

7.3 REFERENCES

8.1 REFERENCES

1.1 BACKGROUND

1.3 REFERENCES

Subject:

2.5 REFERENCES

3.1 INTRODUCTION

3.5 REFERENCES

4.4 REFERENCES

Subject:

6.1 INTRODUCTION

6.5 REFERENCES

7.3 REFERENCES

8.1 REFERENCES

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