Text

License Amendment Request to Amend the Technical Specifications to Revise and Relocate the Reactor Coolant System Pressure and Temperature Limits and Pressurizer Overpressure Protection System Limits to a Pressure and Temperature
	ML22220A248
Person / Time
Site:	Salem
Issue date:	08/07/2022
From:	Sharbaugh D; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML22220A247	List: LR-N22-0012, License Amendment Request to Amend the Technical Specifications to Revise and Relocate the Reactor Coolant System Pressure and Temperature Limits and Pressurizer Overpressure Protection System Limits to a Pressure and Temperature;
References
	LAR S22-01, LR-N22-0012
	Download: ML22220A248 (241)
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Enclosure 7 Contains Proprietary Information to be Withheld from Public Disclosure Pursuant to 10 CFR 2.390 August 7, 2022 10 CFR 50.90 LR-N22-0012 LAR S22-01 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station Unit 2 Renewed Facility Operating License No. DPR-75 NRC Docket No. 50-311

Subject:

License Amendment Request to Amend the Salem Unit 2 Technical Specifications to Revise and Relocate the Reactor Coolant System Pressure and Temperature Limits and Pressurizer Overpressure Protection System Limits to a Pressure and Temperature Limits Report In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) for Salem Generating Station Unit 2 (Salem Unit 2).

The proposed amendment would revise the reactor coolant system pressure-temperature (P-T) limits and relocate the pressurizer overpressure protection system (POPS) enable temperature and lift settings and the P-T limits to a Pressure and Temperature Limits Report (PTLR). provides an evaluation supporting the proposed changes. Attachment 1 of Enclosure 1 provides the existing TS pages marked up to show the proposed changes. Attachment 2 of provides existing TS Bases pages marked up to show the proposed changes and are being provided for information only. Attachment 3 of Enclosure 1 provides additional justification for application of the WCAP-18124-NP-A methodology for the Salem Unit 2 fluence determination.

In addition, WCAP-18124-NP-A, Rev. 0, Supplement 1-NP-A Rev. 0 (ML22153A139) has been approved, which generically approves WCAP-18124-NP-A methodology for use in the extended beltline. provides the Salem Unit 2 Pressure and Temperature Limits Report. contains the non-proprietary Westinghouse analysis (WCAP-16982-NP, Rev. 3) that originally developed the changes to the P-T Limit Curves for license renewal. contains the non-proprietary Westinghouse analysis (WCAP-18571-NP, Rev. 2) that updates the fluence model and validates the changes to the P-T Limit Curves developed for license renewal.

LR-N22-0012 Page2 provides the non-proprietary version of Westinghouse Letter Report, LTR-SCS-20 NP, Revision 1," Salem Unit 2 Pressurizer Overpressure Protection System (POPS) Analysis," (Non-Proprietary Class 3 version with proprietary content redacted) dated July 26, 2021. is an affidavit for withholding Enclosure 7. provides the proprietary version of Westinghouse Letter Report, LTR-SCS-20-52-P, Revision 1, "Salem Unit 2 Pressurizer Overpressure Protection System (POPS) Analysis," dated July 26, 2021. Enclosure 7 contains proprietary information as defined by 10 CFR 2.390. Westinghouse Electric Company LLC (Westinghouse), as the owner of the proprietary information, has executed the affidavit contained in Enclosure 6 pursuant to 10 CFR 20.390(b)(1 ), and requests that the proprietary information in Enclosure 7 be withheld from public disclosure in accordance with 10 CFR 2.390(a)(4).

PSEG requests approval of this LAR in the normal 12 month review period. The Salem Unit 2 reactor vessel will reach 32 Effective Full Power Years by approximately July 15, 2024. The license amendment will be implemented within 60 days of the issue date.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

There are no regulatory commitments contained in this letter.

If you have any questions or require additional information, please contact Mr. Brian Thomas at 856-339-2022.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on Respectfully, 1),JiJJ(

David Sharbaugh Site Vice President Salem Generating Station

Enclosures:

1. Evaluation of Proposed Changes

2. Salem Unit 2 Pressure and Temperature Limits Report (PTLR)

3. Westinghouse Evaluation WCAP-16982-NP, Rev. 3, "Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity," (Non-Proprietary)

4. Westinghouse WCAP-18571-NP, Rev. 2, "Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," (Non-Proprietary)

LR-N22-0012 Page3

5. Westinghouse Letter Report LTR-SCS-20-52-NP, Rev. 1, "Salem Unit 2 Pressurizer Overpressure Protection System (POPS) Analysis," (Non-Proprietary)

6. Affidavit for Withholding

7. Westinghouse Letter Report LTR-SCS-20-52-P, Rev. 1, "Salem Unit 2 Pressurizer Overpressure Protection System (POPS) Analysis," (Proprietary) cc: Administrator, Region I, NRC Project Manager, NRC NRC Senior Resident Inspector, Salem Ms. Ann Pfaff, Manager, NJBNE PSEG Corporate Commitment Tracking Coordinator Site Commitment Tracking Coordinator

LR-N22-0012 LAR S22-01 Enclosure 1 Evaluation of Proposed Changes Table of Contents 1.0

SUMMARY

DESCRIPTION .............................................................................................. 1 2.0 DETAILED DESCRIPTION ............................................................................................... 1 2.1 System Design and Operation .................................................................................. 1 2.2 Current Technical Specification Requirements ......................................................... 1 2.3 Reason for the Proposed Change ............................................................................. 2 2.4 Description of the Technical Specification Changes ................................................. 2 2.5 Description of the Technical Specification Bases Changes ...................................... 3

3.0 TECHNICAL EVALUATION

.............................................................................................. 3 3.1 Background ............................................................................................................... 3 3.2 Technical Analysis .................................................................................................... 4

4.0 REGULATORY EVALUATION

........................................................................................ 12 4.1 Applicable Regulatory Requirements/Criteria ......................................................... 12 4.2 Precedent ................................................................................................................ 13 4.3 No Significant Hazards Consideration .................................................................... 14 4.4 Conclusions............................................................................................................. 16

5.0 ENVIRONMENTAL CONSIDERATION

.......................................................................... 16

6.0 REFERENCES

................................................................................................................ 16 Attachment 1 Mark-up of Proposed Technical Specification Pages Attachment 2 Mark-up of Proposed Technical Specification Bases Pages Attachment 3 Justification for Application of the WCAP-18124-NP-A Methodology for the Salem Unit 2 Fluence Determination

LR-N22-0012 LAR S22-01 1.0

SUMMARY

DESCRIPTION The proposed change revises the Salem Unit 2 Technical Specifications (TS) reactor coolant system pressure-temperature (P-T) limits and relocates the pressurizer overpressure protection system (POPS) enable temperature and lift settings and the P-T limits to a Pressure and Temperature Limits Report (PTLR).

The PTLR contains updates to the P-T limit curves for the beltline regions for the Salem Unit 2 reactor pressure vessel (RPV). The P-T limit curves are developed for 50 effective full power years (EFPY) of operation. The pressure and temperature limits were developed using the KIc methodology detailed in the 1998 through the 2000 Addenda Edition of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G (Reference 1). The methodology is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, (Reference 2) and WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, (Reference 8).

The proposed TS changes are consistent with the guidance of NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," (Reference 3), and with Technical Specification Task Force (TSTF)

Travelers TSTF-419, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR, (Reference 4), as clarified in the NRC's August 4, 2011 letter (Reference 7) and TSTF-233, "Relocate LTOP Arming Temperature to PTLR," (Reference 6).

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Pressure-temperature limit curves have been developed for the RPV in accordance with 10 CFR 50 Appendix G, ASME Code Section XI, Appendix G, and NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, as applicable. These P-T curves account for material property changes due to radiation and ensure that a postulated surface defect having a depth of 1/4 of the RPV material thickness can be safely accommodated in the vessel shell without promoting brittle fracture.

2.2 Current Technical Specifications Requirements TS Limiting Condition for Operation (LCO) 3.4.10.1 requires that Reactor Coolant System (RCS) pressure and temperature be maintained within the pressure and temperature limits specified in TS Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality and inservice leak and hydrostatic testing. The current P-T limit curves in the TSs are applicable to plant operation up to 32 Effective Full Power Years (EFPY). LCO 3.4.10.1 also requires the temperature rate of change to be maintained within specified limits during heatup, cooldown and hydrostatic testing operations above system design pressure.

TS 3.4.10.3 requires the lift setting for the POPS relief valves to be maintained less than or equal to the specified limit when the temperature of one or more of the RCS cold legs is less than or equal to the specified POPS enable temperature.

The POPS enable temperature is also specified in TS 3.1.2.3, 3.4.1.3, 3.4.1.4 and 3.5.3.

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LR-N22-0012 LAR S22-01 2.3 Reason for the Proposed Change The current Unit 2 P-T limit curves expire at 32 EFPY, which Salem Unit 2 is expected to reach by approximately July 15, 2024. The requested approval date for the proposed amendment supports establishment of updated P-T limit curves prior to reaching 32 EFPY.

2.4 Description of the Technical Specification Changes This license amendment request revises the reactor coolant system P-T limit curves (i.e., TS Figures 3.4-2 and 3.4-3) and relocates the POPS enable temperature and lift settings and the P-T limits to a PTLR, as follows:

1. Revise Index to add new PTLR definition 1.20a and update page number for Section 6.9.2.

2. Add a definition in Section 1.0, 1.20a, for the PTLR.

3. Revise TS 3.1.2.3, Charging Pump - Shutdown, # note to change the RCS cold leg temperature from the current value of 312°F to the POPS enable temperature specified in the PTLR.

4. Revise TS 3.4.1.3,

note to change the RCS cold leg temperature from the current value of 312°F to the POPS enable temperature specified in the PTLR.

5. Revise TS 3.4.1.4, ## note to change the RCS cold leg temperature from the current value of 312°F to the POPS enable temperature specified in the PTLR.

6. Revise TS 3.4.10.1 Limiting Condition for Operation, to refer to limits specified in the PTLR.

7. Revise TS 3.4.10.1.a, the maximum heatup rate, to refer to limits specified in the PTLR.

8. Revise TS 3.4.10.1.b, the maximum cooldown rate, to refer to limits specified in the PTLR.

9. Revise TS 3.4.10.1.c, the maximum temperature change, to refer to limits specified in the PTLR.

10. Revise Surveillance Requirements (SR) 4.4.10.1.1 to refer to the limits specified in the PTLR.

11. Revise SR 4.4.10.1.2 to refer to the Figures (P-T Limit Curves) specified in the PTLR.

12. Remove TS Figure 3.4-2, Salem Unit 2 Reactor Coolant System Heatup Limitations.

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LR-N22-0012 LAR S22-01

13. Remove TS Figure 3.4-3, Salem Unit 2 Reactor Coolant System Cooldown Limitations.

14. Revise TS 3.4.10.3 to refer to the POPS enable temperature and relief valve setting specified in the PTLR.

15. Revise TS 3.5.3, # note to change the RCS cold leg temperature from the current value of 312°F to the POPS enable temperature specified in the PTLR.

16. Revise SR 4.5.3.2 to change the RCS cold leg temperature value of 312°F, to refer to the POPS enable temperature specified in the PTLR.

17. Adds a new Specification 6.9.1.11, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to subsection 6.9, Reporting Requirements in Section 6.0, Administrative Controls. The new specification is consistent in format and content with the NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 5.0 and includes:

The individual TSs that address reactor coolant system P-T limits,

References the NRC approved topical report which documents the PTLR methodology, and

Requires the PTLR and any revisions or supplements to be submitted to the NRC.

The marked-up TS pages are provided in Attachment 1.

2.5 Description of Technical Specification Bases Changes The proposed changes to the TS Bases provided in Attachment 2 are provided for information only; changes to the affected TS Bases pages will be incorporated in accordance with TS 6.16, Technical Specifications (TS) Bases Control Program.

3.0 TECHNICAL EVALUATION

3.1 Background

In Salem Unit 2 Amendment 224 (Reference 5), the curves were calculated using the Westinghouse vessel fluence methodology. Allowable pressure-temperature relationships for various heatup and cooldown rates were calculated using methods derived from Appendix G in Section XI of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-14040-NP-A, Rev. 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996, and ASME Boiler and Pressure Vessel Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1," approved March 1999. Adjusted reference temperatures at the nil ductility transition values were developed for the RPV materials in accordance with RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials. The NRC staff approved revising the RPV pressure-temperature limits and extending their validity to 32 effective full-power years.

These are the current P-T limits in the Salem Unit 2 TS.

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LR-N22-0012 LAR S22-01 3.2 Technical Analysis Generic Letter 96-03 provides regulatory guidance regarding relocation of P-T curves and associated numerical limits (such as heatup and cooldown rates) from plant TS to a PTLR (a licensee controlled document). As stated in GL 96-03, a licensee requesting such a change must satisfy the following three criteria:

1. Have NRC-approved methodologies to reference in the TS.

2. Develop a PTLR to contain the P-T limit curves, associated numerical limits, and any necessary explanation, and

3. Modify applicable sections of the TS accordingly.

This License Amendment Request is based on methodologies described in WCAP-14040-A, Rev. 4 with WCAP-18124-NP-A, Rev 0 used as an alternative to WCAP-14040-A, Rev. 4, Section 2.2.

The following table lists the provisions specified for an approved methodology and the information required to be included in the PTLR, as shown in Attachment 1 of GL 96-03 and the corresponding information in the Salem Unit 2 PTLR.

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LR-N22-0012 LAR S22-01 PROVISIONS FOR METHODOLOGY FROM MINIMUM REQUIREMENTS TO PSEG PTLR (Enclosure 2)

ADMINISTRATIVE CONTROLS SECTION IN STS BE INCLUDED IN PTLR COMPLIANCE

1. The methodology shall describe how the neutron fluence is Provide the values of neutron Table 4-1 provides the values of calculated (reference new regulatory guide when it is fluences that are used in the the neutron fluences used in the issued). adjusted reference temperature ART calculation.

(ART) calculation.

2. The Reactor Vessel Material Surveillance Program shall Provide the surveillance capsule Appendix A provides the comply with Appendix H to 10 CFR Part 50. The reactor withdrawal schedule, or reference surveillance capsule withdrawal vessel material irradiation surveillance specimen removal by title and number the schedule.

schedule shall be provided, along with how the specimen documents in which the schedule examinations shall be used to update the PTLR curves. is located.

Reference the surveillance Appendix A also contains the capsule reports by title and Westinghouse surveillance number if ARTs are calculated capsule reports of the four using surveillance data. capsules withdrawn to date.

3. Low temperature overpressure protection (LTOP) system Provide setpoint curves or Section 4.14 provides the LTOP limits developed using NRC- approved methodologies may setpoint values system setpoint value.

be included in the PTLR.

4. The adjusted reference temperature (ART) for each reactor Identify both the limiting ART Table 4-6 provides the limiting beltline material shall be calculated, accounting for values and limiting materials at ART values and the limiting irradiation embrittlement, in accordance with Regulatory the 1/4t and 3/4t locations (t = materials.

Guide 1.99, Revision 2. vessel beltline thickness).

PWRs - identify RTPTS value in Section 4.11 references the accordance with 10 CFR 50.61. location of the RTPTS values in the Westinghouse analysis (Enclosure 4).

5. The limiting ART shall be incorporated into the calculation Provide the P/T curves for heatup, Figures 3-1 and 3-2 provide the of the pressure and temperature limit curves in accordance cooldown, criticality, and P/T curves, including heatup, with NUREG- 0800, SRP Section 5.3.2, Pressure- hydrostatic and leak tests. cooldown, criticality, and Temperature Limits. hydrostatic and leak tests.

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LR-N22-0012 LAR S22-01 PROVISIONS FOR METHODOLOGY FROM MINIMUM REQUIREMENTS TO PSEG PTLR (Enclosure 2)

ADMINISTRATIVE CONTROLS SECTION IN STS BE INCLUDED IN PTLR COMPLIANCE

6. The minimum temperature requirements of Appendix G to Identify minimum temperatures on Figures 3-1 and 3-2 P/T curve 10 CFR Part 50 shall be incorporated into the pressure and the P/T curves such as minimum annotations provide the temperature limit curves. boltup temperature and hydrotest minimum boltup and hydrotest temperature. (hydrostatic) temperatures.

7. Licensees who have removed two or more capsules should Provide supplemental data and Sections 4.5 and 4.6 provide compare for each surveillance material the measured calculations of the chemistry descriptions of how the data increase in reference temperature (RTNDT) to the predicted factor in the PTLR if the from multiple surveillance increase in RTNDT; where the predicted increase in RTNDT is surveillance data are used in the capsules are used in the based on the mean shift in RTNDT plus the two standard ART calculation. chemistry factor and ART deviation value (2) specified in Regulatory Guide 1.99, calculation.

Revision 2. If the measured value exceeds the predicted value (increase in RTNDT + 2), the licensee should provide Evaluate the surveillance data to Appendix C contains the a supplement to the PTLR to demonstrate how the results determine if they meet the credibility evaluation of the affect the approved methodology. credibility criteria in Regulatory surveillance data.

Guide 1.99, Revision 2.

Provide the results.

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LR-N22-0012 LAR S22-01 Revised pressure-temperature curves were developed for hydrostatic pressure and leak tests, core not critical, and core critical conditions. The PTLR containing the P-T limit curves, associated tabulated numerical limits, and necessary explanation, is provided in Enclosure 2. The revised curves have been developed for 50 EFPY.

The P-T limit curves were generated through End of License Extension (EOLE), i.e., 60 years of operation, using the KIc methodology detailed in the 1998 through the 2000 Addenda Edition of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G. The P-T limit curve generation methodology is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4.

The NRC concluded WCAP-14040-A, Revision 4 was acceptable for referencing as a PTLR methodology, subject to three conditions:

a. Licensees who wish to use WCAP-14040, Revision 3, as their PTLR methodology must provide additional information to address the methodology requirements discussed in provision 2 in the table of Attachment 1 to GL 96-03 related to the RPV material surveillance program.

The methodology requirements discussed in provision 2 in the table of Attachment 1 to GL 96-03 require that licensees briefly describe the surveillance program. Licensees should identify by title and number the report containing the Reactor Vessel Surveillance Program and surveillance capsule reports. The RPV surveillance program was verified to comply with the Requirements of 10 CFR 50, Appendix H in Section 3.0.3.2.9, Reactor Vessel Surveillance, of the Salem Units 1 and 2 license renewal safety evaluation report (SER) (Reference 11). The requirement to adhere to 10 CFR 50, Appendix H is retained in TS SR 4.4.10.1.2. The PTLR specifically identifies:

The surveillance programs compliance with 10 CFR 50, Appendix H,

That the surveillance capsule withdrawal schedule is contained in UFSAR Section 5.2.4.4, and

A complete listing of the Reactor Vessel Surveillance Program and surveillance capsule reports

b. Contrary to the information in WCAP-14040, Revision 3, licensee use of the provisions of ASME Code Cases N-588, N-640, or N-641 in conjunction with the basic methodology in WCAP-14040, Revision 3, does not require an exemption since the provisions of these Code Cases are contained in the edition and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a. When published, the approved revision (Revision 4) of TR WCAP-14040 should be modified to reflect this NRC staff conclusion.

This condition was satisfied upon publication of WCAP-14040-A, Revision 4.

c. As stated in WCAP-14040, Revision 3, until Appendix G to 10 CFR Part 50 is revised to modify/eliminate the existing RPV flange minimum temperature requirements or an exemption request to modify/eliminate these requirements is approved by the NRC for a specific facility, the stated minimum temperature must be incorporated into a facility's P-T limit curves.

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LR-N22-0012 LAR S22-01 The Salem Unit 2 P-T limit curves were developed with the flange requirements of 10 CFR 50, Appendix G.

The Westinghouse evaluation supporting the P-T Limit Curves, WCAP-18571-NP, Rev. 2, Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, (Reference 9) is contained in this submittal as Enclosure 3. This report verifies the P-T Limit Curves that were generated in Westinghouse evaluation WCAP-16982-NP, Rev. 3 for 50 EFPY (Reference 10) developed for license renewal remain applicable for Salem Unit 2.

Refer to Enclosure 4 for this report. The reports contents are summarized below.

Section WCAP-16982-NP, Rev. 3 WCAP-18571-NP, Rev. 2 1.0 Identifies the RPV analyses which Describes the evaluation to are considered Time-Limited confirm the applicability of the P-T Aging Analyses for License curves from WCAP-16982-NP Renewal. with consideration of updated fluence values.

2.0 References the methodology for Provides the methodology for the the fluence projections (use of fluence projections (use of 3-DORT discrete ordinates code dimensional RAPTOR-M3G), as Version 3.2 and the BUGLE-96 well as the results of the cross-section library), as well as projections, including fluence at the results of the projections. the primary nozzle forgings.

3.0 Evaluates the fracture toughness Evaluates the fracture toughness properties of the RPV beltline and properties of the RPV beltline and extended beltline materials, extended beltline materials, including a summary of the Best- including a summary of the Best-Estimate Chemistry and Initial Estimate Chemistry and Initial RTNDT Values for the Salem Unit 2 RTNDT Values for the Salem Unit RPV materials. 2 RPV materials (including nozzle forgings).

4.0 Summarizes the Pressurized Discusses the results of the Thermal Shock (PTS) results. Salem Unit 2 surveillance capsule program, as well as Salem Unit 2 applicable sister plant surveillance capsule data.

5.0 Provides the Upper Shelf Energy Describes the calculation of (USE) values at 50 EFPY with a Chemistry Factors using comparison to the screening Regulatory Guide 1.99, Revision criteria. 2 for the Salem Unit 2 RPV materials and its sister plant's materials.

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LR-N22-0012 LAR S22-01 Section WCAP-16982-NP, Rev. 3 WCAP-18571-NP, Rev. 2 6.0 Explains the detailed Explains the Adjusted Reference methodology for developing the Temperature (ART) calculation Salem Unit 2 P-T Limit Curves, methodology and results at the including requirements to address 1/4T and 3/4T RPV locations the metal temperature of the through 50 EFPY, including the closure head/vessel flange and determination of the limiting ART the minimum boltup temperature. values Also discusses the minimum LTOP enable temperature.

Provides the tabulated results supporting the Salem Unit 2 P-T Limit Curves, as well as the generated curves.

7.0 Provides the surveillance capsule Provides the basis for P-T Curve withdrawal schedule, including applicability for those P-T Curves the recommendation for the next developed in WCAP-16982-NP, capsule withdrawal. Rev. 3 (refer to Section 6.0 of that report).

8.0 Determines the emergency Lists References for the report.

response guideline (ERG) P-T Limit category.

9.0 Lists References for the report. Not Applicable to this Report.

Appendix A Provides Stress Intensity Factors Describes the evaluation of the (KIT) at 1/4T and 3/4T locations credibility of the Salem Unit 2 for 50 EFPY. surveillance capsule program data using the criteria provided in Regulatory Guide 1.99, Revision 2.

Appendix B Not Applicable to this Report. Contains the Pressurized Thermal Shock (PTS) evaluation as required by 10 CFR 50.61 and ERG classification.

Appendix C Not Applicable to this Report. Validation of the radiation transport models based on neutron dosimetry measurements Neutron Fluence Calculations:

The methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET." The neutron transport evaluation methodology described in the WCAP report is based on the guidance of Regulatory Guide 1.190. The NRC concluded the calculational methodology described in WCAP-18124-NP-A, Revision 0 is acceptable for use in calculating RPV neutron fluence provided the following limitations and conditions are met:

1. Applicability of WCAP-18124-NP, Revision 0, is limited to the reactor pressure vessel region near the active height of the core based on the uncertainty analysis performed and the measurement data provided. Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (for example, pressure-temperature limits, material stress and strain), margin 9

LR-N22-0012 LAR S22-01 assessment, or a combination thereof, for applications of the method to components including, but not limited to, the reactor pressure vessel upper circumferential weld, and reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

The methodology described in WCAP-18124-NP-A, Revision 0 was applied to Salem Unit 2 extended beltline materials and RCS inlet and outlet nozzles. Additional justification for the Salem Unit 2 application is provided in Attachment 3 to this enclosure (Reference 13). Based on the fluence uncertainty analysis, benchmarking, and margin assessment described in Attachment 3, the applicability of the RAPTOR-M3G fluence determination methodology is justified for the Salem Unit 2 RPV extended beltline region and RCS inlet and outlet nozzle fluence determination. In addition, WCAP-18124-NP-A, Rev. 0, Supplement 1-NP-A Rev. 0 (ML22153A139) has been approved, which generically approves WCAP-18124-NP-A methodology for use in the extended beltline.

2. Least squares adjustment is acceptable if the adjustments to the measured or calculated ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the inconsistency should be disqualified.

This limitation applies in situations where the least squares analysis is used to adjust the calculated values of neutron exposure. For Salem Unit 2, Limitation and Condition 2 does not apply as the least-squares procedures were not used to adjust the calculated fast neutron (E > 1.0 MeV) fluence values for RPV materials evaluated in the updated reactor vessel integrity analysis (Enclosure 4). The least-squares results were only used to compare the calculations and measurements from the evaluated dosimetry and validate the neutron transport models, and those comparisons showed satisfactory results.

A validation of the fluence model is provided in Appendix C of Enclosure 4.

The fluence is based upon operation for 50 EFPY. Cycle-specific calculations were performed for Cycles 1 through 24. Note that future fluence projection data beyond Cycle 24 are based on Cycles 20 and 23.

The calculated fast neutron fluences at the end of plant life (50 EFPY) are provided below:

Parameter Fluence (n/cm2)

Peak Surface (Intermediate Shell Plates) 2.05x1019 Peak 1/4 T (Intermediate Shell Plates) 1.22x1019 Limiting Beltline Material Peak Surface 1.47x1019 (Intermediate Shell Longitudinal Weld Seams 2-442 B & C)

Limiting Beltline Material Peak 1/4 T 0.876x1019 (Intermediate Shell Longitudinal Seams 2-442 B & C)

Use of Salem Unit 2 and Sister Plant Surveillance Capsule Results and Adjusted Reference Temperature:

As discussed in Appendix C of the PTLR, the surveillance data for intermediate shell plate B4712-2 (Salem Unit 2 surveillance program) and lower shell longitudinal weld seams 3-442A, B, and C (sister plant Diablo Canyon Unit 2 surveillance program) are deemed credible.

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LR-N22-0012 LAR S22-01 As reported in WCAP-18571-NP, Rev. 2, the Diablo Canyon Unit 2 surveillance data for Heat #

21935/12008 were deemed credible, hence the calculated lower Adjusted Reference Temperature (ART) of Salem Unit 2 lower shell longitudinal weld seams 3-442A, B, and C calculated as determined by Regulatory Guide 1.99, Rev. 2 Position 2.1 supersedes those calculated with Regulatory Guide 1.99, Rev, 2, Position 1.1. Therefore, the limiting material for Salem Unit 2 are intermediate shell longitudinal weld seams 2-442B and C.

This Diablo Canyon Unit 2 surveillance program is applicable to Salem Unit 2 Lower Shell Longitudinal Weld Seams 3-442 A, B, and C since both the surveillance weld and reactor vessel welds were fabricated with wire Heat # 21935/12008 with Linde 1092 flux. The surveillance data is reported in WCAP-17315-NP, Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations, and were adjusted to account for difference in weld chemistry and irradiation temperature. This approach is similar to that taken for the McGuire Units 1 and 2 Measurement Uncertainty Recapture (MUR) Power Uprate, evaluated in WCAP-17455-NP, McGuire Units 1 and 2 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations and approved in the SER (ML13073A041),

which also used Diablo Canyon Unit 2 surveillance data to evaluate a McGuire Unit 1 reactor vessel weld with wire Heat # 21935/12008.

Applicability of WCAP-16982-NP P-T Limit Curves:

Instead of re-creating the P-T curves in WCAP-18571-NP, Rev. 2, the P-T curves in WCAP-16982-NP, Rev. 3 were evaluated for applicability for 50 EFPY. Below is a summary of the Adjusted Reference Temperatures (ART) needed for this applicability evaluation (re-produced from Table 7-1 of WCAP-18571-NP, Rev. 2).

Summary of the Limiting ART Values 1/4T Limiting ART (°F) 3/4T Limiting ART (°F)

WCAP-16982-NP, Rev. 3 WCAP-18571-NP, Rev. 2 WCAP-16982-NP, Rev. 3 WCAP-18571-NP, Rev. 2 209 191.6 150 138.1 The above table shows that the limiting ART values at the 1/4T and 3/4T locations will not exceed the ART values used in the P-T limit curves of WCAP-16982-NP. Furthermore, the limiting initial RTNDT value of 15°F for the closure head flange and the vessel flange remains appropriate, and the P-T limit curves flange notch and bolt-up temperature require no change or further consideration. Therefore, the EOLE (50 EFPY) P-T limit curves generated in WCAP-16982-NP can be implemented in the PTLR.

Pressure-Temperature Curve Evaluation:

The beltline region of the RPV was evaluated to develop the 50 EFPY P-T curves (Reference 10),

which were re-validated in Reference 9. The WCAP-104040-A, Rev. 4 methodology used to generate the P-T curves in this submittal is approved by the NRC. When re-validating the P-T Curves in Reference 9, the credible surveillance capsule data were incorporated into the ART values in accordance with RG 1.99, Rev. 2.

The 50 EFPY P-T curves ensure that adequate RPV safety margins against non-ductile failure will continue to be maintained during normal operations, anticipated operational occurrences, and inservice leak and hydrostatic testing. Together, these measures ensure that the integrity of the Reactor Coolant System will be maintained for the life of the plant.

11

LR-N22-0012 LAR S22-01 Proposed revisions to applicable sections of the TS have been prepared and are provided in to this submittal. These proposed changes are consistent with the guidance provided in GL 96-03, as supplemented by TSTF-419 and TSTF-233.

Pressurizer Overpressure Protection System (POPS) Power Operated Relief Valves (PORVs) setpoint selection The Low Temperature Overpressure Protection System (LTOPS), known as the Pressurizer Overpressure Protection System (POPS) at Salem, provides RCS pressure relief capability to mitigate the overpressure transients that may occur during cold shutdown, heatup, and cooldown operations to minimize the potential for challenging reactor vessel integrity limits when operating at low temperature conditions (i.e., 10 CFR 50, Appendix G limits). At Salem, the pressurizer Power Operated Relief Valves (PORVs), with reduced lift settings, provide a method of overpressure protection. The POPS PORV setpoints are selected in accordance with the NRC approved methodology in WCAP-14040-A, Rev. 4, such that the peak pressure during the design basis Mass Injection (MI) and Heat Injection (HI) transients at low RCS temperature conditions (i.e., 350°F) will not exceed the isothermal Appendix G Pressure-Temperature (P-T) limits.

Updated P-T limits have been developed for Salem Unit 2 through a service life of 50 EFPY. As part of updating the P-T limits, an analysis was performed (Reference 12) to validate the applicability of the POPS analysis previously performed (Reference 14) to ensure that the POPS configuration, relief valve setpoints, and operating limitations protect the revised P-T limits at 50 EFPY.

The following acceptance criteria were used to determine the POPS PORV setpoints:

1. The peak RCS pressure resulting from the design basis MI and HI transients shall not exceed the minimum of the steady-state adjusted Appendix G limits and the PORV piping limit.

2. The minimum RCS pressure resulting from the design basis MI and HI transients should not drop below the RCP No. 1 Seal P limit.

The maximum allowable PORV setpoint is determined based on the adjusted Appendix G limit or the PORV piping limit, whichever is more limiting. The adjusted Appendix G limit is the Appendix G limit minus the transmitter P and wide range pressure instrument uncertainty. The maximum allowable PORV setpoints for the MI and HI transients were determined as a function of indicated RCS temperature, where the final maximum allowable PORV setpoint would bound both the MI and HI transient maximum allowable PORV setpoints.

The maximum allowable Salem Unit 2 POPS PORV setting was determined by the analysis to be 434 psig. The current Salem Unit 2 POPS PORV setting of 375 psig will be documented in the PTLR to maintain conservative margin.

The POPS enable temperature of 292°F with temperature uncertainty applied was determined by Westinghouse Letter Report LTR-SCS-20-52 (Reference 12) using ASME code case N-641.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36, "Technical specifications," provides the regulatory requirements for the content required in the TSs which includes limiting conditions for operation (LCO's), surveillance requirements and administrative controls. Previously the plant-specific P-T limits had been 12

LR-N22-0012 LAR S22-01 incorporated into the TS and controls were placed on operation and testing by the associated specification. This proposed change revises the TS to relocate the P-T limit curves to a licensee controlled document in accordance with the guidance of Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," TSTF-419, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR, and TSTF-233, Relocate LTOP Arming Temperature to PTLR.

10 CFR 50.60 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to 10 CFR 50.

Appendix G is the regulatory basis for P-T curves for light water reactors. Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Appendix G also requires that the reference temperature and Charpy upper-shelf energy for reactor vessel beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel.

The NRC has established requirements in 10 CFR 50, Appendix G, "Fracture Toughness Requirements," in order to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. Appendix G requires that the P-T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods and margins of safety of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code were used to generate the P-T limits. 10 CFR Part 50, Appendix G contains additional requirements for the metal temperature of the closure head flange and vessel flange regions or when the core is critical. Also, Appendix G requires that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Appendix H to 10 CFR Part 50 provides requirements related to facility RPV material surveillance programs. Salem Unit 2 demonstrates its compliance with the requirements of 10 CFR 50, Appendix H through withdrawal and analysis of surveillance capsules in accordance with the schedule contained in the UFSAR.

Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.

4.2 Precedents The NRC has approved similar license amendments to relocate the P-T limit curves to a PTLR.

Examples include:

Salem Generating Station, Unit 1, (License Amendment No. 339 issued by NRC letter dated October 8, 2021, ADAMS Accession No. ML21230A018)

Hope Creek Generating Station, (License Amendment No. 209 issued by NRC letter dated December 14, 2017, ADAMS Accession No. ML17324A840)

Palo Verde Nuclear Generating Station (License Amendment No. 178 issued by NRC letter dated February 25, 2010, ADAMS Accession No. ML100480188)

Sequoyah Nuclear Plant Units 1 and 2, (License Amendment Nos. 294/284 issued by NRC 13

LR-N22-0012 LAR S22-01 letter dated September 15, 2004, ADAMS Accession No. ML042600465)

Prairie Island Nuclear Generating Plant Units 1 and 2, (License Amendment Nos. 135/127 issued by NRC letter dated May 4, 1998, ADAMS Accession No. ML022260558)

Beaver Valley Power Station Units 1 and 2, (License Amendment Nos. 256/138 issued by NRC letter dated July 15, 2003, Accession No. ML031960399)

Byron/Braidwood, (License Amendment Nos. 98 (Byron) and 89 (Braidwood) issued by NRC letter dated January 23, 1998, ADAMS Accession No. ML020860605)

The NRC previously approved changes similar to the proposed pressure and temperature limit change in this License Amendment Request for other nuclear power plants. Examples include:

Arkansas Nuclear One Unit 2, (License Amendment 311 issued by NRC letter dated November 27, 2018, ADAMS Accession No. ML18298A012)

Catawba Nuclear Station Units 1 and 2, (License Amendment Nos. 306/302 issued by NRC letter dated August 4, 2020, ADAMS Accession No. ML20174A045)

R. E. Ginna Nuclear Power Plant, (License Amendment No. 106 issued by NRC letter dated February 23, 2009, ADAMS Accession No. ML083530806)

H. B. Robinson Steam Electric Plant Unit No. 2, (License Amendment No. 248 issued by NRC letter dated November 22, 2016, ADAMS Accession No. ML16285A404)

Indian Point Generating Unit No. 3, (License Amendment No. 258 issued by NRC letter dated September 3, 2015, ADAMS Accession No. ML15226A159)

Seabrook Station Unit No. 1, (License Amendment No. 151 issued by NRC letter dated November 2, 2015, ADAMS Accession No. ML15096A255)

Joseph M. Farley Nuclear Plant Units 1 and 2, (Amendment Nos.193/189 issued by NRC letter dated October 2, 2013, ADAMS Accession No. ML13249A386) 4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) requests an amendment to Renewed Facility Operating License No. DPR-75 for Salem Generating Station Unit 2 (Salem Unit 2). The proposed Technical Specification (TS) changes modify the Salem Unit 2 TS by revising the pressure and temperature (P-T) limits and relocate the pressurizer overpressure protection system (POPS) enable temperature and lift settings and the P-T limits to a Pressure and Temperature Limits Report (PTLR).

PSEG has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

14

LR-N22-0012 LAR S22-01

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change revises the pressure and temperature (P-T) limits and relocates the pressurizer overpressure protection system (POPS) enable temperature and lift settings and the P-T limits to a Pressure and Temperature Limits Report (PTLR). The P-T limit curve generation methodology is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4 with WCAP-18124-NP-A used as an alternative to WCAP-14040-A, Revision 4, Section 2.2.

10 CFR 50 Appendix G establishes requirements to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants.

Implementing this NRC approved methodology does not reduce the ability to protect the RCPB as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components.

Incorporation of the new methodology for calculating P-T curves, and the relocation of the P-T curves, POPS Enable Temperature, and POPS limits from the TS to the PTLR provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions.

The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The change in methodology for calculating P-T limits and the relocation of those limits to the PTLR do not alter or involve any design basis accident initiators. RCPB integrity will continue to be maintained in accordance with 10 CFR 50 Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), and the installed equipment is not being operated in a new or different manner.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed changes involve a significant reduction in a margin of safety?

Response: No 15

LR-N22-0012 LAR S22-01 The proposed changes do not affect the function of the RCPB or its response during plant transients. Calculating the Salem Unit 2 P-T limits using the NRC approved methodology ensures adequate margins of safety relating to RCPB integrity are maintained. The proposed changes do not alter the manner in which the Limiting Conditions for Operation P- T limits for the RCPB are determined. The operability requirements for equipment assumed to operate for accident mitigation are not affected.

Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.

Based upon the above, PSEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions Therefore, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. 1998 through the 2000 Addenda Edition of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G

2. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, dated May, 2004 (ADAMS Accession No. ML050120209)

3. NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," dated January 31, 1996 (ADAMS Accession No. 9601290350)

4. TSTF-419, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR," dated August 4, 2003 (ADAMS Accession No. ML012690234)

5. Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendment Re:

Increase Licensed Power Levels From 3,411 MWT to 3,459 MWT, dated May 25, 2001 16

LR-N22-0012 LAR S22-01 (ADAMS Accession No. ML011350051)

6. TSTF-233, Relocate LTOP Arming Temperature to PTLR, dated July 16, 1998

7. NRC Letter, Implementation of Travelers TSTF-363, Revision 0, Revise Topical Report References In ITS 5.6.5, COLR [Core Operating Limits Report], TSTF-408, Revision 1, Relocation of LTOP [Low Temperature Overpressure Protection] Enable Temperature and PORV [Power-Operated Relief Valve] Lift Setting To the PTLR [Pressure-Temperature Limits Report], and TSTF-419, Revision 0, Revise PTLR Definition and References In ISTS [Improved Standard Technical Specification] 5.6.6, RCS [Reactor Coolant System] PTLR, dated August 4, 2011 (ADAMS Accession No. ML110660285)

8. WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET," dated July, 2018 (ADAMS Accession No. ML18204A010)

9. WCAP-18571-NP, Revision 2, Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, dated July 2022

10. WCAP-16982-NP, Revision 3, Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity, dated December 2020

11. Safety Evaluation Report Related to the License Renewal of Salem Nuclear Generating Station, Units 1 and 2, June 13, 2011 (ADAMS Accession No. ML11164A051)

12. Westinghouse Letter Report LTR-SCS-20-52-P, Rev. 1, Salem Unit 2 Low Temperature Overpressure Protection System (LTOPS) / Pressurizer Overpressure Protection System (POPS) Analysis dated July 26, 2021

13. Westinghouse Letter PNJ-REAC-TM-AA-000001, Rev. 1, dated July 6, 2022, Justification of Using RAPTOR-M3G for Reactor Pressure Vessel Extended Beltline Materials at Salem Unit 2

14. Westinghouse Letter PSE-09-18, Revision 0, Transmittal of LTOPS Setpoint Evaluation for Salem Units 1 and 2, March 2009 17

LR-N22-0012 LAR S22-01 Attachment 1 Mark-up of Proposed Technical Specification Changes The following Technical Specification pages for Renewed Facility Operating License DPR-75 are affected by this change request:

I XVIII 1-5 3/4 1-9 3/4 4-3 3/4 4-4a 3/4 4-27 3/4 4-28 3/4 4-29 3/4 4-31 3/4 5-7 3/4 5-8 6-24b 6-24c (new page)

LR-N22-0012 LAR S22-01 INSERT 1 (Page 1-5)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.20a The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, and the Overpressure Protection System setpoint and enable temperature, for the current reactor vessel fluence period. The pressure and temperature limits shall be determined for each fluence period in accordance with Technical Specification Section 6.9.1.11.

INSERT 2 (Page 6-24b) 6.9.1.11 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, POPS enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.1.2.3, "Charging Pump-Shutdown"

2. Specification 3.4.1.3, Reactor Coolant System Hot Shutdown

3. Specification 3.4.1.4, Reactor Coolant System Cold Shutdown"

4. Specification 3/4.4.10.1, "Reactor Coolant System Pressure/Temperature Limits"

5. Specification 3.4.10.3, "Reactor Coolant System Overpressure Protection Systems"

6. Specification 3/4.5.3, Emergency Core Cooling Systems ECCS Subsystems - Tavg <

350°F

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC.

1. WCAP-14040-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May, 2004.

2. WCAP-18124-NP-A, Rev 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018, may be used as an alternative to Section 2.2 of WCAP-14040-A Rev. 4.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplements thereto.

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS ........ .. .. .. ................................................................................. 1-1 ACTION.... ....... ........... .. .. .. ................................................................................. 1-1 AXIAL FLUX DIFFERENCE .. ................................................................................. 1-1 CHANNEL CALIBRATION . .. ................................................................................. 1-1 CHANNEL CHECK ...... .. .. .. ................................................................................. 1-1 CHANNEL FUNCTIONAL TEST ............................................................................. 1-1 CONTAINMENT INTEGRITY ................................................................................. 1-2 CORE ALTERATION ... .. .. .. ................................................................................. 1-2 CORE OPERATING LIMITS REPORT ................................................................... 1-2 DOSE EQUIVALENT I-131 .. ................................................................................. 1-2 DOSE EQUIVALENT XE-133 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME ......................................... 1-3 FREQUENCY NOTATION . .. ................................................................................. 1-3 FULLY WITHDRAWN .. .. .. .. ................................................................................. 1-3 GASEOUS RADWASTE TREATMENT SYSTEM .................................................. 1-3 IDENTIFIED LEAKAGE .. .. .. ................................................................................. 1-3 INSERVICE TESTING PROGRAM......................................................................... 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM) ............................................. 1-4 OPERABLE - OPERABILITY ................................................................................. 1-4 OPERATIONAL MODE - MODE ............................................................................. 1-4 PHYSICS TESTS ......... .. .. .. ................................................................................. 1-5 PRESSURE BOUNDARY LEAKAGE ..................................................................... 1-5 PROCESS CONTROL PROGRAM (PCP) .............................................................. 1-5 PURGE-PURGING ...... .. .. .. ................................................................................. 1-5 QUADRANT POWER TILT RATIO ......................................................................... 1-5 RATED THERMAL POWER . ................................................................................. 1-5 REACTOR TRIP SYSTEM RESPONSE TIME ....................................................... 1-6 REPORTABLE EVENT .. .. .. ................................................................................. 1-6 SHUTDOWN MARGIN. .. .. .. ................................................................................. 1-6 SOLIDIFICATION......... .. .. .. ................................................................................. 1-6 SOURCE CHECK ........ .. .. .. ................................................................................. 1-6 STAGGERED TEST BASIS .. ................................................................................. 1-6 THERMAL POWER ..... .. .. .. ................................................................................. 1-7 UNIDENTIFIED LEAKAGE .. ................................................................................. 1-7 VENTILATION EXHAUST TREATMENT SYSTEM ................................................ 1-7 VENTING . ....... ........... .. .. .. ................................................................................. 1-7 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) ................ 1-5 SALEM - UNIT 2 I Amendment No. 323

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY ................................................................................................................................................6-1 6.2 ORGANIZATION Onsite and Office Organizations .............................................................................................6-1 Facility Staff ................................................................................................................................................6-1 Shift Technical Advisor .....................................................................................................................6-7 6.3 FACILITY STAFF QUALIFICATIONS ...................................................................................................6-7

6. 4

DELETED .....................................................................................................................................................................6-7 6.5 REVIEW AND AUDIT (THIS SECTION DELETED) .....................................................................6-8 6.6 REPORTABLE EVENT ACTION .....................................................................................................................6-16 6.7 SAFETY LIMIT VIOLATION ........................................................................................................................6-16 6.8 PROCEDURES AND PROGRAMS .....................................................................................................................6-17 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ....................................................................................................................................6-20 c 6.9.2 SPECIAL REPORTS ....................................................................................................................................6-24b 6.10 RECORD RETENTION ..........................................................................................................................................6-25 6.11 RADIATION PROTECTION PROGRAM ......................................................................................................6-26 6.12 HIGH RADIATION AREA .................................................................................................................................6-27 6 .13 PROCESS CONTROL PROGRAM (PCP) ...................................................................................................6-28 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) ........................................................................6-29 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS .......................................................................................6-2 9 6.16 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM ....................................6-30 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM ............................................................6-31 SALEM - UNIT 2 XVIII Amendment No. 298

DEFINITIONS PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the Updated FSAR, 2) authorized under the provisions of 10CFR50.59, or 3) otherwise by the Commission.

Insert 1 PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.

SALEM - UNIT 2 1-5 Amendment No. 262

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE.'

APPLICABILITY: MODES 4, 5 and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one charging pump is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required b y the INSERVICE TESTING PROGRAM.

the POPS enable temperature specified in the PTLR,

A maximum of one centrifugal charging pump shall be OPERABLE while in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312° F, MODE 5, or MODE 6 when the head is on the reactor vessel.

SALEM - UNIT 2 3/4 1-9 Amendment No. 300

REACTOR COOLANT SYSTEM HOT SHtJ'l'DOWN LIMITING CONDITION FOR OPERATION 3.4.l.3 a. At least two of the coolant loops listed below shall be OPERABLE:

l. Reactor Coolant Loop 21 and its associated steam generator and reactor coolant pump,*

2. Reactor Coolant Loop 22 and its associated steam generator and reactor coolant pump,*

3. Reactor Coolant Loop 23 and its associated steam generator and reactor coolant pump,*

4. Reactor Coolant Loop 24 and its associated steam generator and reactor coolant pump,*

5. Residual Heat Removal Loop 2l,

6. Residual Heat Removal Loop 22.

b At least one of the above coolant loops shall be in operation.**

APPLICABILITY: MODB 4.

ACTION:

a. With less than the above required loops OPBRABLB, immediately initiate corrective action to return the required loops to OPBRABLB status as soon as-possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and iDmediately initiate corrective action'to the POPS enable temperature return the required cooluit loop to operation.

specified in the PTLR

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperature less than or equal to 312 ° P unless 1) the pressurizer water volume is less than 1650 cubic feet {equivalent to approximately 92%

of level) or 2) the secondary water temperature of each steam generator is less than 50°P ahova each of the RCS cold leg temperatures. I

- All reactor coolant pumps and residual heat removal pumps may be de energized for up to l hour provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°7 below saturation temperature.

I SALEM - ONIT 2 3/4 4-3 Amendment No. 206

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat removal loops shall be OPERABLE* and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5.##

ACTION:

a. With less than the above required loops operable, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no RHR loop in operation; suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

4.4.1.4 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

One RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation. Additionally, four filled reactor coolant loops, with at least two steam generators with their secondary side water levels greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop. the POPS enable temperature specified in the PTLR

1. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 312°F unless 1) the pressurizer water volume is less than 1650 cubic feet (equivalent to approximately 92% of level), or 2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

Systems supporting RHR loop operability may be excepted as follows:

a. The normal or emergency power source may be inoperable.

- The residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.

SALEM - UNIT 2 3/44-4a Amendment No. 282

REACTOR COOLANT SYSTEM 3/4.4.10 PRESSURE/TEMPERATURE LIMITS limits specified in the PTLR with:

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100°F in anyone hour period, rate within the limits specified in the PTLR,

b. A maximum cooldown of 100°F in anyone hour period, and

c. A maximum temperature change of less than or equal to 5°F in anyone hour period, during hydrostatic testing operations above system design pressure.

APPLICABILITY: At all times.

within the limits specified in the ACTION: PTLR With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T avg and pressure to less than 200"F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

specified in the PTLR 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

the P-T Limit Curves specified in the PTLR.

SALEM - UNIT 2 3/44-27 Amendment No. 282

DELETE THIS FIGURE AND REPLACE WITH THIS PAGE LEFT INTENTIONALLY BLANK Limiting Material Property Weld 3*442 A&C Initial RT NoT -ss*c RT NOT after 32 EFPY: 1/4T 199'F 3/4T 140'F 21500 I

22110 Lellk T - Limit I""

2000 I I I

17110 JUnna1pblbleJ Ope- *ar:upe.b .. 1 v

OperMian I

1 v

I ~

"-- C-Limlt 100 .,.._ l'lltr 7110

,100

,H - pl'lltr

.,.._ -,t-1100

/ C.......,. L i m i t - on lc.. lnMiwlce lrydrwbtda teet

- ( 2 7 W I') for-MNicepwlodupiiDIZIII'PV l::.:.Detl*""'l .

250 -pT-0 I

0 110 100 1110 200 250 300 3110 400 450 1100 11110 Moderator Temperatura (Deg. F)

Figure 3.4-2 SALEM UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 32 EFPY WITH MAXIMUM HEATUP RATE OF 100 F/HR. CURVE CONTAINS MARGIN FOR INSTRUMENT ERRORS SALEM - UNIT 2 Amendment No. 224 3/4 4-28

DELETE THIS FIGURE AND REPLACE WITH THIS PAGE LEFT INTENTIONALLY BLANK Limiting Material Property Weld 3-442 A&C Initial RT NOT -56'C RTNor after 32 EFPY: 1/4T 199'F 3/4T 140'F 2500 2250 2000 1750 6'

&1500

I I I I

11250 11000 tl 1-1;-

ll'lllr I:..

750 500

~

_,_..,.. v 250 r- j.;;"" ... ~ J 0 I I I 0 so 100 150 200 2SO 300 350 400 450 soo sso Modenotor Tempenotu111 (Deg, F)

Figure 3.4-3 SALEM UNIT 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE FOR THE FIRST 32 EFPY WITH MAXIMUM COOLDOWN RATE OF 100 F/HR. CURVE CONTAINS MARGIN FOR INSTRUMENT ERRORS.

SALEM - UNIT 2 3/4 4-29 Amendment No. 224

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.10.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two Pressurizer Overpressure Protection System relief valves (POPS) with a lift setting of less than or equal to 375 psig, or the value specified in the PTLR,

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 3.14 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 3l2°F, except when the reactor vessel head is removed.

ACTION: the POPS enable temperature specified in the PTLR,

a. With one POPS inoperable in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to 312°F, restore the inoperable POPS to OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent(s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

b. With one POPS inoperable in MODES 5 or 6 with the Reactor Vessel Head installed, restore the inoperable POPS to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or complete depressurization and venting of the RCS through at least a 3.14 square inch vent(s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE fltatus.

c. With both POPSs inoperable, depressurize and vent the RCS through a 3.14 square inch vent(s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

d. In the event either the POPS or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the POPS or vent(s) on the transient and any corrective action necessary to prevent recurrence.

e. LCD 3.0.4.b is not applicable when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.10.3.1 Each POPS shall be demonstrated OPERABLE by:

SALEM - UNIT 2 3/4 4-31 Amendment No. 258

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tavg <350"F LIMITING CONDITION FOR OPERATION

==============:====================================================

3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump and associated flow path capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and;

1. Discharging into each Reactor Coolant System (RCS) cold leg.

b. One OPERABLE residual heat removal pump and associated residual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation and;

1. Discharging into each RCS cold leg, and; upon manual initiation,

2. Discharging into two RCS hot legs.

APPLICABILITY: MODE 4.

ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no l"CCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Tavg lessthan 350°F by use of alternate heat removal methods.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

d. LCO 3.0.4.b is not applicable to ECCS high head subsystem

A maximum of one safety injection pump or one centrifugal charging pump shall be OPERABLE in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312°F, Mode 5, or Mode 6 when the head is on the reactor vessel.

the POPS enable temperature specified in the PTLR, SALEM - UNIT 2 3/4 5-7 Amendment No. 258

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tavg < 350°F 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per applicable Surveillance Requirements of 4.5.2.

4.5.3.2 All safety injection pumps and centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated to be inoperable in accordance with the Surveillance Frequency Control Program while in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to 312°F, MODE 5, or MODE 6 when the head is on the reactor vessel by either of the following methods: the POPS enable temperature specified in the PTLR,

a. By verifying that the motor circuit breakers have been removed from their electrical power supply circuits or,

b. By verifying that the pump is in a recirculation flow path and that two independent means of preventing RCS injection are utilized.

SALEM - UNIT 2 3/4 5-8 Amendment No. 282

ADMINISTRATIVE CONTROLS

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment;

4. The number of tubes plugged during the inspection outage; and

d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results.

e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;

f. The results of any SG secondary side inspections.

Insert 2 SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.

6.9.3 DELETED 6.9.4 When a report is required by ACTION 1, 4, 8 OR 9 of Table 3.3-11 "Accident Monitoring Instrumentation", a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.

Move to new page 6-24c SALEM - UNIT 2 6-24b Amendment No. 320

LR-N22-0012 LAR S22-01 Attachment 2 Mark-up of Proposed Technical Specifications Bases Changes (For Information Only)

Proposed Technical Specifications Bases Revised Pages B3/4 4-1 B3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-10 B 3/4 4-11 B 3/4 4-12 B3/4 4-13 B3/4 4-14 B3/4 4-15 B3/4 4-16 B3/4 4-17 B3/4 5-2

LR-N22-0012 LAR S22-01 INSERT 1 (Bases Page 3/4 4-7)

In order to provide sufficient safety margins for protection against non-ductile failure, the pressure/temperature (P/T) limit curves for heatup and cooldown, inservice leak and hydrostatic testing, and criticality are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the 1/4T and 3/4T locations of the beltline and extended beltline materials for the applicability term stated in the PTLR. The selection of such a limiting RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

INSERT 2 (Bases Page 3/4 4-8)

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in UFSAR Section 5.2. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNDT. To ensure that the radiation embrittlement effects are accounted for in the calculation of the P/T limit curves, the most limiting RTNDT includes the radiation induced shift, RTNDT corresponding to the end of the period for which curves are generated. This adjusted reference temperature (ART) can be predicted based upon the fluence and the copper and nickel content of the material in question. The ART is based upon the largest value of RTNDT computed by the methodology presented in Regulatory Guide 1.99, Revision 2. The PTLR contains the results of the ART evaluations. The predicted neutron fluence, as a function of Effective Full Power Years (EFPY), has been calculated based on the guidance of Regulatory Guide 1.190.

The PTLR contains the results of the fluence evaluations.

INSERT 3 (Bases Page 3/4 4-11)

The heatup and cooldown curves are composite curves, constructed based on a point-by-point comparison of the steady-state and finite rate data, as well as the reactor vessel and head flange, to identify the most restrictive curve. The use of the composite curve is necessary to set conservative heatup/cooldown limitations because at any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations have more restrictive limits, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limit curve requirement is that it be 40°F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for inservice leak rate testing.

l/i ,t JW.c::rOR COOI,N!T IXITII BASIS 3/4

4. l RP,c;TOJt CQOLAlff LOOPS NfP C:OQLANT qRc::pµTIQN The plaz,t ia deaig,oed to operate with all reactor coolaz,t loop* in operation, maet th* DNB deisin criteria du.-in; all no.,...l operation* az,d anticipated traz,aianta. In KODZS 1 and 2 with l*** thaz, all coolaz,t loop in operation, thi* apecification requir** that th* plant he in at l*a*t BOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In KODZ 3, a *in;l* reactor coolaz,t loop provide aufficient heat rU10val for rU10vin; decay heat1 but, *in;le failure coniderationa require all l oop he in operation whenever th* rod control syatem ia aner;ized and at l*aat one loop he in operation when the rod control syatem i* deanar;ized.

In KODZ 4, a ain;l* reactor coolaz,t loop or RBJI loop provide aufficient heat rUIQval for rUIQvin; decay heat; hut, in;l* failure conaiderationa require that at leat 2 loop* be OPBR.AIILll. Thu, if the reactor coolant loops are not OPBR.AIILll, thi *pecification require that two RBR loop ha OPJIUl!LE.

In MODS 5, in;la failure conaidaration* require that two REil loop h*

OPSR.AIILll. Per *upport ayat..,.., Service Water (SW) and Component Coolin; (CC),

component redundancy ia n*c****ry to *n*ur* no aingle active component failure will cau** th* lo** of Dacey Beat Removal. On* pipin; path of SW and CC i*

adequate when it aupporta both RBR loop*. Th* aupport Y*t..,.. needed before entering into tbe deair*d confii'\lZ'&tion (e.g., one **rvic* water loop out for m&intenance in Mod** 5 az,d 6) are controlled by procadur*** and include th*

fol loving:

A requirement that two RBJI, two CC and two SW pwszpa, powered from two different vital bu*** b* kept operable

A liting of th* ective (air/actor operated) valv** in th* affected flow path to be locked open or diaabled Note that four filled reactor coolant l oop,-with at leaat two team generator* with at leaat their 1econdary_ aid* water level greater than or equal to 5\ (narrow range), may b* a@tituted for one reaidual beat removal loop. Thi* eoaur** t.hat a ai.D;l* failure doe not cauae a lo** of decay beat r-=c:,val.

The operation of on* Jteactor Coolaz,t Pump or one RBJI Pump provid**

adequate flow to enure ai.x.i..Dg, preve.nt atratification and produce gradual reactivity cbLJlg...--duri.Dg aoron concentration reduction. in t.h* Reactor Coolant Syt.,.. Th* reactivity chan;* ret* aeociated with Boron concentration reduetioa.a will, therefore, be withic th* capability of operator recogoitioc a.cd control. the POPS enable temperature specified in the PTLR The r*etrictiona on at&rtin; a Jteactor Coolant Pum;, below P-7 with one or

""'r* acii cold l*;* l*** than or equal to 12*, a.-e provided to pr**ant JtCS pr*ure trLJli*nta, cauaed by ecerff addition* froa t.he aecocdary syate:m, wich could exceed th* li.&it* of Appendix Oto 10 Part 50. Th* acs will be protected agaiot overpr***ur* tra.nienta and will not axce*d th* liait* of Appendix G by either (l) reatrictin; th* water volume in th* pr***urizer

!thereby providing a velum* into which th* primary coolant can .xpand, or 121 by re*tricting the atartin; of Reactor Coolant Pump* to tho** tiJu* whan ecoodary water te.m:perature i.J:i each eteaa generator i* l*** than so*r above each of th* acs cold le; tu,peratur***

S .u.BX - tllUT 2 a 3/4 4*1 Amend!Mnt No. 197

REACTOR COOLANT SYSTEM BASES 3/4.4.10 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G.

those specified in the the PTLR

1) The reactor coolant temperature and pressure and system heatup and cooldown rate (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon.

therein.

a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit specified in the lines shown. Limit lines for cooldown rates between those PTLR. presented may be obtained by interpolation.

The limits specified in the PTLR b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of nonductile failure only. For normal operation, other inherent non-ductile plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can The limit lines be achieved over certain pressure-temperature ranges.

specified in the PTLR

2) These limit lines shall be calculated periodically using methods provided below.

in Technical Specification 6.9.1.11.

3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70°F.

4) The pressurizer heatup and cooldown rates shall not exceed 100°F/hr and 200°F/hr, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 32o*F.

5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM ElBS-82, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1996 Summer Addenda to Section XI of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-14040-NP-A, Rev. 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", January 1996, and ASME Boiler and Pressure Vessel Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", approved March 1999.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT~, at the end of 32 effective full power years of service life. The 32 EFPY service life period is chosen such that the limiting RTN~ at the l/4T location in the core region is greater than the RTNoT of the limiting unirradiated material. The selection of such a limiting RTNM assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

Replace with Bases Insert 1 SALEM - UNIT 2 B 3/4 4-7 Amendment No. 224

REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been tested to determine their initial RTN~;

the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTmiT. An adjusted reference temperature, (ART), based upon the fluence and the copper and nickel content of the material in question, can be predicted.

The ART is based upon the largest value of RTN~ computed by the methodology presented in Regulatory Guide 1.99, Revision 2. The ART for each material is given by the following expression:

ART - Initial RTN~ + ~RTN~ + Margin Initial RTNm is the reference temperature for the unirradiated material.

&RTNoT is the mean value of the adjustment in reference temperature caused by the irradiation and is calculated as follows:

~RT.~ - Chemistry Factor x Fluence Factor The Chemistry Factor, CF (F), is a function of copper and nickel content. It is given in Table 83/4.4-2 for welds and in Table 83/4.4-3 for base metal (plates and forgings) . Linear interpolation is permitted.

The predicted neutron fluence as a function of Effective Full Power Years (EFPY) has been calculated and is shown in Figure 83/4.4-1. The fluence factor can be calculated by using Figure 83/4.4-2. Also, the neutron fluence at any depth in the vessel wall is determined as follows:

-0.24X f ~ (f surface) x (e )

where "f surface" is from Figure 83/4.4-1, and X (in inches) is the depth into I._

the vessel wall.

Finally, the "Margin" is the quantity in *F that is to be added to obtain conservative, upper-bound values of adjusted reference temperature for the calculations required by Appendix G to 10 CFR 50.

If a measured value of initial RTN~ for the material in question is used, a 1 may be taken as zero. If generic value of initial RT~ is used, a, should be obtained from the same set of data. The standard deviations, for ~RT*~*

a~, are 28°F for welds and l7°F for base metal, except that ad need not exceed 0.50 times the mean value of ~RTNM surface.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNcT at the end of 32 EFPY.

Replace with Bases Insert 2 SALEM - UNIT 2 B 3/4 4-8 Amendment No. 224

REACTOR COOLANT SYSTEM BASES Values of ~RT*~ determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM El85, are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR Part 50, Appendix H. The heatup and cooldown curves must be recalculated when the ~RT*~ determined from the surveillance capsule exceeds the calculated &RTNor for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section XI of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-14040-NP-A, Rev. 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", January 1996, and ASME Boiler and Pressure Vessel Code Case N-640, ~Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", approved March 1999.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section XI as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure.

To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTN~' is used and this includes the radiation induced shift, ~RT*~ corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1 , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. K1c is obtained from the reference fracture toughness curve, defined in ASME Code Case N-640. The K1c curve is given by the equation:

Krc = 33.2 + 20.734 exp [0.02(T-RTN~)] (1) where Krc is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RTNoT* Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

(2)

SALEM - UNIT 2 B 3/4 4-9 Amendment No. 224

REACTOR COOLANT SYSTEM Page intentionally left blank BASES where KrM is the stress intensity factor caused by membrane (pressure) stress.

Krr is the stress intensity factor caused by the thermal gradients.

Krc is provided by the code as a function of temperature relative to the RTNor of the material.

C 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, Krc is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTN 01 , and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Krr* for the reference flaw are computed. From Equation (2) the pressure stress intensity factors are obtained and from these the allowable pressures are calculated.

COOL DOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.

Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the l/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the aT developed during cooldown results in a higher value of Krc at the 1/4T location for finite cooldown rates than for steady-state operation.

Furthermore, if conditions exist such that the increase in K1c exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the l/4T location, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

SALEM - UNIT 2 B 3/4 4-10 Amendment No. 224

REACTOR COOLANT SYSTEM BASES HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite. heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stress at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature. Therefore, the Krc for the 1/4T crack during heatup is lower than the Krc for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K1cs for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the l/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Replace with Bases Insert 3 SALEM - UNIT 2 8 3/4 4-11 Amendment No. 224

REACTOR COOLANT SYSTEM BASES Finally, the new 10CFR50 rule which addresses the metal temperature of the closure head flange regions is considered. This 10CFR50 rule states that the metal temperature of the closure flange regions must exceed the material RTNoT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Salem}. Table B3/4.4-l indicates that the limiting RTNoT of 2B"F occurs in the closure head flange of Salem Unit 2, and the minimum allowable temperature of this region is 14B"F at pressures greater than 621 psig. These limits do not affect Figures 3.4-2 and 3.4-3.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

valves the POPS enable temperature specified in the PTLR.

The OPERABILITY of two POPSs or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 312"F. Either POPS has valve adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1} the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50°F above the RCS cold leg temperatures, or (2} the start of an Intermediate Head Safety Injection pump and its injection into a water solid RCS, or the start of a High Head Safety Injection pump in conjunction with a running Positive Displacement pump and its injection into a water solid RCS. The minimum electrical power sources required to assure POPS operability (based on POPS meeting the single failure criteria}

consist of a normal (via offsite power) and an emergency (via batteries) power source for each train of POPS. Emergency diesel generators are not required for POPS to meet single failure criteria and therefore are not required for POPS OPERABILITY.

POPS LCO 3.0.4.b is not applicable to an inoperable LTOP system when entering MODE 4.

There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable LTOP system. The provisions of LCO 3.0.4.b, which allow entry into a MODE or other specitied condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

SALEM - UNIT 2 B 3/4 4-12 Amendment No. 256

TABLF. 1/4.4-1 SALEM UNIT 2 REACT( ESSEL TOUGHNESS DATA Delete Table and mark page "Page left intentionally blank" Average Upper Shell Energy Component Plate No. Material Cu (\) Ni T 50 ft-lb RT (oF) Normal to Principal or Weld Type (\) (oF) 35 - Mil Principal Working Working No. Temp Direction Direction (oF) lft-lb) (ft-lb)

Closure Hd Dome 84708 A533BCLl 0.11 0.70 -40 45* -15* 82.5 127 Closure Hd Peel 85007-3 A533BCL1 0.12 0.57 -20 15* -20* 97* 149 Closure Hd Peel B4707-l A533BCL1 0.10 0.55 0 51* o* 84* 129 Closure Hd Peel 84707-3 A533BCL1 0.13 0.63 0 66* 6* 84* 129.5 Closure Hd Flng B4 702-1 A508CL2 - 0.68 28* 39* 28* 104* 160 Vessel Flange B5001 AS08CL2 - 0.70 12* 4* 12* 107* 164 Inlet Nozzle 84703-1 A508CL2 - 0.69 60* 62* 60* >72* >111**

Inlet Nozzle B4703-2 A508CL2 - 0.69 60* 25* 60* >61* >94**

Inlet Nozzle B4703-3 A508CL2 - 0.68 60* 32* 60* >71* >109**

Inlet Nozzle B4703-4 A508CL2 - 0.81 60* 40* 60* so* 123.5 Outlet Nozzle B4704-1 A508CL2 - 0.84 60* a* 60* 82* 126 Outlet Nozzle B4704-2 A508CL2 - 0.77 60* 20* 60* 75* 116 Outlet Nozzle B4704-3 A508CL2 - 0.69 28* a* 28* 82* 126 Outlet Nozzle B4704-4 A508CL2 - 0.71 60* 40* 60* 77* 119 Upper Shell B4711-1 A533BCL1 0.11 0.55 0* 50* 0* 87* 134 Upper Shell B4711-2 A533BCL1 0.14 0.56 -10 60* 0* 79* 122 Upper Shell B4711-3 A533BCL1 0.12 0.58 -10 88* 28* 69* 107 I

Inter. Shell B4712-1 A533BCL1 0.13 0.56 0 <60 0 106 138 Inter. Shell B4712-2 A533BCL1 0.12 o;62 20 72 12 97 127.5 Inter. Shell B4712-3 A533BCL1 0.11 0.57 -so 70 10 107 116 Lower Shell B4713-l A5338CL1 0.12 0.60 10 68 8 98 127 Lower Shell 84713-2 A5338CL1 0.12 0.57 -20 68 8 103 135.5 Lower Shell B4713-3 A5338CL1 0.12 0.58 -10 70 10 121 135.5 Bottom Hd Peel B4709 1 A5338CL1 0.12 0.60 -30 54* -6* 90* 139 Bottom Hd Peel B4709-2 A5338CL1 0.12 0.58 -20 42* -18* 89* 137.5 Bottom Hd Peel B4709-3 A533BCL1 0.11 0.56 -20 71* 11* 93* 143 Bottom Head B4710 A533BCL1 0.12 0.60 -30 60* 0* 77* 118 Circum. Weld Bet 8-442 - 0.28 0.74 - - -56*** - -

Nozzle Shell &

Int. Shell Circum. Weld Bet 9-442 - 0.197 0.060 - - -56*** 99.7 -

Int. Shell & Lower Shell Int. Shell 2-442 - 0.219 0.735 - - -56*** 96.2 -

Vertical Weld [A,B,C]

Lower Shell 3-442 - 0.213 0.867 - - -56*** 114 -

Vertical Weld [A,B,C]

Est>mated per NRC Standard Rev>ew Plan Sect>on 5.8.2.

- 100% Shear not reached

- - Estimate per Pressurized Thermal Shock Rule, 10 CFR 50.61 SALEM - UNIT 2 8 3/4 4-13 Amendment No. 224

TAIILE a J/4.4.2

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0 :ao :ao :ao 10 :ao :ao :ao 0.01 .:ao :ao :ao 10 :ao :ao :ao 0.01 21 21 rt rt rt rt rt o.os 22 u 41 41 41 41 41 u u w w w w w 0.04 0.01 21 0.01 17 _12 12 12 12 12 0.07 32 II II 16 16 16 16 0.01 31 II 10 101 101 101 101 0.011 .0 11 M 111 122 122 122 0.10 II 122 133 lU lJI 0.11 101 130 lU 141 141 0.12 12 72 lOS lJI lU 111 111 O.lS II 11 101 131 112 172 111 0.14 11 101 142 111 112 1a 0.11 112 141 111 111 :100 0.11 70 111 141 111 111 211 0.17 11 12 111 151 114 'Ill 221 0.11 11 16 122 1U 111 214 230 0.11 u 100 121 157 111 2:10 231 o.:ao 104 121 110 1M 223 241 0.21 12 101 133 1M 111 221 252 0.22 11 112 131 111 :100 m 257 0.23 101 117 1.0 lit 301 231 313 o.u 101 121 1U 113 :aoe 2Jt 211 0.25 110 12t 141 111 2U 212 0.21 us 130 111 110 212 Ul rtl o.rr 111 1M 111 1M 211 Ul 210 0.21 122 131 110 111 211 251 214 0.21 121 142 1W 111 217 222 2W 0.10 111 141 111 1M 225 257 210 0.31 lJI 111 172 1a *221 210 213 0.32 1.0 111 111 231 313 211 0.33 1U 110 110 2M 211 211 O.M o.u "'

1U 1M 111 1M 111 2011 212 2JI Ul 211 m

302 305 O.JI 111 172 111 211 Ul rtl 301 0.31 1U 117 111 2:10 Ul rtl Sll 0.31 lH 1U :100 223 210 211 314 0.31 171 111 301 227 2N 211 317 o..o 175 111 'Ill 231 251 211 ~~

SALEM UliiT Z B J/4 4-14. Amendment No. 86

TABLE B 314.4-3 Delete Table and mark page "Page left intentionally blank" CBiamY r&C!DI POl IASI III!AI.,

21"bW\-I c.,.,

0 u . . 2:12 1..22 1....i2 0 30 30 30 30 30 30 30 0.01 30 30 30 30 30 30 30 0.01 30 30 30 30 30 30 30 0.03 30 30 30 30 30 30 30 0.04 22 21 O.CII 11 11 31 31 31 31 0.01 21 ~ ~ ~ ~ -~ ~

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" "11 "11 "11 "11 11 0.08 ~ 13 5I 5I II 5I II 0.10 41 5I II II 11 11 11 0.11 45 12 72 74 T1 T1 71 0.12 48 17 78 13 0.13 13 71 II 11 0.14 57 15 81 100 101 101 101 0.15 11 10 110 111 117 117 0.11 II 14 104 111 123 121 125 0.17 u II 110 127 132 135 135 0.11 73 82 111 134 141 14.4 14.4 0.18 71 87 130 142 110 114 114 0.20 12 102 1111 148 118 154 111 0.21 II 107 121 111 111 172 174 0.22 81 112 134 181 171 111 114 0.23 117 131 117 114 180 184 o.a.. 100 121 142 112 181 188 :104 0.21 104 121 141 171 188 301 214 0.21 108 130 151 110 306 211 221 0.21 114 134 111 114 211 221 230 0.21 118 131 1110 111 211 233 238 0.21 124 142 114 181 221 a.l.1 241 0.30 121 141 117 184 221 a..e 217 0.31 134 111 172 1M 221 211 211 0.32 131 111 171 :1102 231 2110 214 0.31 14.4 110 110 301 314 214 :an 0.34 148 114 114 301 231 211 280 0.31 113 111 111 212 241 212 211 0.31 151 173 181 211 341 211 301 0.~ 112 111 1M 230 341 211 301 0.31 1M 112 300 223 210 211 313 0.38 171 111 liOS 221 2W 211 317 0.40 175 1H 'liJ'I 231 217 m SALEM rJliiT 2 B 3/4 4-15 Amendment No. 86

Delete Figure and mark page "Page left intentionally blank" 1.0E+20 This Curve Represents the Fluence at the Inner Radius of the Limiting Long~udinal Weld Seam Located at the 30° Azimuth N"

E 1.0E+19

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!:o Fluence at Vessel Inner Radius

I Ill z 30° Azimuth 1.0E+18 1.0E+17 0 5 10 15 20 25 30 35 Service Life (Effective Full Power Years)

Figure B 3/4.4-1 Fast neutron f1uence (E > 1 MeV) as a function of full power service life (EFPY)

SALEM - UNIT 2 B 3/4 4-16 Amendment No. 224

Delete Figure and mark page "Page left intentionally blank" Best Available Copy

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SALEM - UNIT 2 s 3/4 4-17 Amendment No. 26

EMERGENCY CORE COOLING SYSTEMS BAES ECCS SUBSYSTEMS (C<:mtinued)

With the RCS temperature below 350 ° F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one safety injection pump or one centrifugal charging pump to be OPERABLE and the Surveillance requirement to verify all safety injection pumps except the allowed OPERABLE safety injection pump to be inoperable below 312 ° F provides assurance that a mass addition pressure transient can be relieved by the operation of a single POPS relief valve.

the POPs enable temperature specified in the PTLR When rimning a safety injection pump with the RCS temperature less than 312 °F with the potential for injecting into the RCS and creating a mass addition pressure transient, two independent means of preventing reactor coolant system injection will be utilized, The two independent means can be satisfied by any of the following methods:

(1) A manual isolation valve locked in the closed position; or (2) Two mant1al isolation valves closed; or (3) One motor operated valve closed and its breaker de-energized and control circuit fuses removed; or (4) One air' operated valve closed and its air supply maintained in such a manner as to ensure that the valve will remain closed.

The surveillance requirements, which are provided to ensure the OPERABILITY of each component, ensure that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. The safety analyses make the assumptions with respect to: 1) both the maximum and minimum total system resistance, and 2) both the maximum and minimum branch injection line resistance. These resistances, in conjunction with the ranges of potential pump performance, are used to calculate the maximum and minimum ECCS flow assumed in the safety analyses.

The maximum and minimum flow surveillance requirements in conjunction with the mximum and minimum pump performance curves ensures that the assumptions of total system resistance and the distribution of that system resistance among the various paths are met.

The maximum total pump flow surveillance requirements ensure the pump runout limits of 560 gpm for the centrifugal charging pumps and 675 gpm for the safety injection pumps are not exceeded. Due to the effect of pump sµction boost alignment, the runout limits for the surveillance criteria are s 554 gpm for C/SI pumps, s 664 gpm for SI pumps in cold leg alignment and s 654 gpm for SI pumps in hot leg alignment.

The surveillance requirement for the maximum difference between the maximum and minimum individual injection line flows ensure that the minimum individual.injection line resistance assumed for the spilling line following a LOCA is met.

LCO 3.0.4.b is not applicable to an inoperable ECCS high head subsystem when entering MODE 4. There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable ECCS high head subsystem.

The provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

SALEM - UNIT 2 B 3/4 5-2 Amendment No. 258

LR-N22-0012 LAR S22-01 Attachment 3 Justification for Application of the WCAP-18124-NP-A Methodology for the Salem Unit 2 Fluence Determination

LR-N22-0012 LAR S22-01 1.0 Introduction RAPTOR-M3G has been used in the fast neutron (E > 1.0 MeV) fluence evaluation in support of updating reactor pressure vessel fluence for Salem Unit 2 to generate normal operation heatup and cooldown limit curves. Per the requirement of Regulatory Issue Summary (RIS) 2014-11 (Reference 1), the reactor pressure vessel (RPV) nozzles and any RPV materials that exceed a fast neutron (E > 1.0 MeV) fluence of 1E+17 n/cm2 at the end-of-license extension (EOLE) must be evaluated with respect to fracture toughness. The RPV materials evaluated for Salem Unit 2 are listed in Table 1. Certain materials in this table were determined to have fluence values less than 1E+17 n/cm2 and therefore did not require specific evaluation with respect to neutron embrittlement and fracture toughness.

The active core height extends from -182.88 cm to +182.88 cm in the RAPTOR-M3G model that represents the axial extension of the traditional beltline region of the reactor vessel which, by definition, is the region that directly surrounds the effective height of the active core (Reference 2). When compared to the axial elevations of the RPV materials evaluated in Table 1, the intermediate shell, intermediate shell longitudinal welds, intermediate shell to lower shell circumferential weld, lower shell, and lower shell longitudinal welds are subsumed under the traditional beltline region. Therefore, these materials have been approved by the United States Nuclear Regulatory Commission (USNRC) for generic application of RAPTOR-M3G for fast neutron (E > 1.0 MeV) fluence determination per WCAP-18124-NP-A (Reference 3).

The 50 effective full power years (EFPY) fast neutron (E > 1.0 MeV) fluence at the EOLE and locations for the RPV materials in the extended beltline region for Salem Unit 2 are listed in Table 2. The limitations and conditions for using RAPTOR-M3G for fast neutron (E > 1.0 MeV) fluence determination are stipulated in Section 4.0 of the safety evaluation letter captured in Reference 3. The limitations and conditions of Reference 3 are quoted below:

1. Applicability of WCAP-18124-NP, Revision 0, is limited to the RPV region near the active height of the core based on the uncertainty analysis performed and the measurement data provided. Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (e.g.,

pressure- temperature limits, material stress/strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the RPV upper circumferential weld and reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

2. Least squares adjustment is acceptable if the adjustments to the M/C ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the inconsistency should be disqualified.

The second limitation and condition listed above does not apply as the least-squares procedures were not used to adjust the calculated fast neutron (E > 1.0 MeV) fluence values for RPV materials evaluated in the reactor vessel integrity analysis. The least-squares results were only used to compare the calculations and measurements from the evaluated dosimetry and validate the neutron transport models, and those comparisons showed satisfactory results.

1 of 17

LR-N22-0012 LAR S22-01 The Reference 3 neutron fluence methodology, however, has been used to determine the fast neutron (E > 1.0 MeV) fluence values for RPV materials in the extended beltline region.

Therefore, additional justification should be provided for use of the methodology in this region.

The following information provides the additional justification of using RAPTOR-M3G for fast neutron (E > 1.0 MeV) fluence determination for the RPV extended beltline regions at Salem Unit 2 by summarizing the additional benchmark data from a Westinghouse 4-loop pressurized water reactor (PWR). In addition, WCAP-18124-NP-A, Rev. 0, Supplement 1-NP-A Rev. 0 (ML22153A139) has been approved, which generically approves WCAP-18124-NP-A methodology for use in the extended beltline.

2 of 17

LR-N22-0012 LAR S22-01 Table 1: Locations for Reactor Vessel Materials Azimuthal Axial Location[2]

Location[1] [°] [cm]

Material Min Max Min Max Outlet Nozzle to Nozzle Shell Welds[3]

- Nozzle 1 22.0 270.12

- Nozzle 2 158.0 270.12

- Nozzle 3 202.0 270.12

- Nozzle 4 338.0 270.12 Inlet Nozzle to Nozzle Shell Welds[3]

- Nozzle 1 67.0 265.04

- Nozzle 2 113.0 265.04

- Nozzle 3 247.0 265.04

- Nozzle 4 293.0 265.04 Nozzle Shell

- Plate 1 300.0 60.0 233.53 485.07

- Plate 2 60.0 180.0 233.53 485.07

- Plate 3 180.0 300.0 233.53 485.07 Nozzle Shell Longitudinal Welds[3]

- Weld 1 60.0 233.53 485.07

- Weld 2 180.0 233.53 485.07

- Weld 3 360.0 233.53 485.07 Nozzle Shell to Intermediate Shell Circumferential Weld - Centerline[3] 0.0 360.0 233.53 Intermediate Shell

- Plate 1 0.0 120.0 -42.93 233.53

- Plate 2 120.0 240.0 -42.93 233.53

- Plate 3 240.0 360.0 -42.93 233.53 Intermediate Shell Longitudinal Welds[3]

- Weld 1 0.0 -42.93 233.53

- Weld 2 120.0 -42.93 233.53

- Weld 3 240.0 -42.93 233.53 Intermediate Shell to Lower Shell Circumferential Weld - Centerline[3] 0.0 360.0 -42.93 Lower Shell

- Plate 1 300.0 60.0 -313.52 -42.93

- Plate 2 60.0 180.0 -313.52 -42.93

- Plate 3 180.0 300.0 -313.52 -42.93 Lower Shell Longitudinal Welds[3]

- Weld 1 60.0 -313.52 -42.93

- Weld 2 180.0 -313.52 -42.93

- Weld 3 360.0 -313.52 -42.93 Lower Shell to Lower Vessel Head Circumferential Weld - Centerline[3] 0.0 360.0 -313.52 Notes:

1. Azimuthal locations are indexed to = 0.0 as shown on reactor pressure vessel general assembly drawing.

2. Axial elevations are indexed to Z = 0.0 at the midplane of the active fuel stack.

3. Longitudinal welds are modelled to 3.49 cm width. Circumferential welds are modelled using a 3.18 cm axial width.

3 of 17

LR-N22-0012 LAR S22-01 Table 2: Fast Neutron (E > 1.0 MeV) Fluence at 50 EFPY and Locations for RPV Extended Beltline Region Materials Axial Unit 2 Location[1] [n/cm2]

Material [cm]

Outlet Nozzle to Nozzle Shell 270.12 1.78E+16 Welds - Lowest Extent Inlet Nozzle to Nozzle Shell Welds 265.04 2.26E+16

- Lowest Extent Nozzle Shell 233.53 2.44E+17 Nozzle Shell Longitudinal Welds 233.53 1.83E+17 Nozzle Shell to Intermediate Shell 233.53 2.80E+17 Circumferential Weld[2]

Lower Shell to Lower Vessel Head

-313.52 3.24E+15 Circumferential Weld[2]

Notes:

1. Values listed are indexed to Z = 0.0 at the midplane of the active fuel stack and only the closest distance to the core midplane is listed.

2. The fluence values for these welds are reported at the maximum location within the width.

2.0 Additional Benchmarking Measurements In order to collect measurement benchmark data for the extended beltline region, three sets of ex-vessel neutron dosimetry (EVND) have been installed at the elevation of the reactor vessel support for a Westinghouse 4-loop plant. The elevation of the reactor vessel support is approximately 8.5 feet above the core midplane. The specific axial locations of the EVND capsules to the core midplane (Z = 0.0 cm) and times of irradiation are listed in Table 3. The dosimeter foils included in these EVND capsules are listed in Table 4. The measured dosimetry reactions for those foils are listed in Table 5.

4 of 17

LR-N22-0012 LAR S22-01 Table 3: Location and Time of Irradiation for Sensor Sets Analyzed at RPV Supports Sensor Azimuthal Axial Elevation Cycle(s) of Capsule ID Location Location [cm] Irradiation E Ex-Vessel 180° 257.99 11 A Ex-Vessel 225° 255.75 11 K Ex-Vessel 180° 257.99 12 - 19 M Ex-Vessel 180° 257.99 20 R Ex-Vessel 135° 261.65 20 Table 4: Foil Sensor Set Contents in EVND at RPV Supports Capsule ID Radiometric Monitor Foils Fe Ni Cu Ti Co Nb U-238 Np-237 E x x X x x x x A x x X x x x x K x x X x x x M x x X x x x R x x X x x x Table 5: Measured Dosimetry Reactions in EVND at RPV Supports Neutron Energy Product Half-Material Reaction of Interest Response(1) Life(2)

Copper 63Cu (n,) 60Co 4.53-11.0 MeV 5.271 y Titanium 46Ti (n,p) 46Sc 3.70-9.43 MeV 83.788 d Iron 54Fe (n,p) 54Mn 2.27-7.54 MeV 312.13 d Nickel 58Ni (n,p) 58Co 1.98-7.51 MeV 70.86 d 238U 238U (n,f) 137Cs 1.44-6.69 MeV 30.05 y Niobium 93Nb (n,n) 93mNb 0.95-5.79 MeV 16.13 y 237Np 237Np (n,f) 137Cs 0.68-5.61 MeV 30.05 y Cobalt - Al 59Co (n,g) 60Co Thermal 5.271 y Note(s):

(1) Energies between which 90% of activity is produced (235U fission spectrum). Ref. ASTM E844-18.

(2) Half-life data is from ASTM E1005-16.

2.1 Additional Benchmarking Neutron Transport Calculations The neutron transport calculations for the additional benchmarking at the 4-loop Westinghouse plant extended beltline region followed the Westinghouse fluence methodology described in Reference 3, which is the same methodology used for the neutron transport calculations performed in support of Salem Unit 2 reactor pressure vessel fluence evaluation work. In the application of this methodology to the fast neutron exposure evaluations for the 4-loop Westinghouse plant EVND dosimetry sets at the RPV supports, forward transport calculations were carried out to directly solve for the space- and energy-dependent scalar flux, (r,,z,E).

5 of 17

LR-N22-0012 LAR S22-01 For the additional benchmark analysis, all of the transport calculations were carried out using the RAPTOR-M3G three-dimensional discrete ordinates code and the BUGLE-96 (Reference 4) cross-section library. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray group cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modelled with an S16 order of angular quadrature.

A plan view of the reactor model is shown in Figure 1. In addition to the core, reactor internals, RPV, and concrete bioshield, the model also included explicit representations of the surveillance capsules, RPV clad, and RPV nozzles and supports. Section views of the reactor model are shown in Figure 2 and Figure 3.

In developing the model of the reactor geometry, nominal design dimensions were used for the various structural components. Water temperatures (and densities) in the core, bypass, and downcomer regions of the reactor were taken to be representative of full-power operating conditions. These coolant temperatures were varied on a cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, etc.

The r,,z geometric mesh description of the reactor model consisted of 233 radial by 186 azimuthal by 469 axial mesh intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion used in the calculations was 0.001.

The core power distributions used in the additional benchmarking transport analysis included fuel-assembly-specific initial enrichments, burnups, and axial power distributions. This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of the fuel cycle-averaged neutron fluence rate, which, when multiplied by the appropriate fuel cycle length, provide the incremental fast neutron (E > 1.0 MeV) fluence exposure for each fuel cycle. The energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial 235U enrichment and burnup history of the individual fuel assemblies. From the assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined. These fuel-assembly-specific neutron source strengths derived from the detailed isotopics were then converted from fuel pin Cartesian coordinates to the r,,z spatial mesh arrays used in the RAPTOR-M3G discrete ordinates calculations.

6 of 17

LR-N22-0012 LAR S22-01 Figu re 1 : Reactor Geometry - Plan View at Core Mid plane 7 of 17

LR-N22-0012 LAR S22-01 F ig u re 2: Reactor Geometry - Section View at Outlet Nozzle Centerline 8 of 17

LR-N22-0012 LAR S22-01 Figure 3: Reactor Geometry - Section View at Inlet Nozzle Centerline 9 of 17

LR-N22-0012 LAR S22-01 2.2 Additional Benchmarking Dosimetry Evaluations The evaluations of the neutron sensor sets contained in the EVND dosimetry capsules at the 4-loop Westinghouse plant RPV supports followed the state-of-the-art least-squares dosimetry evaluation methodology described in Section 3.0 of Reference 3.

Least-squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best-estimates for key exposure parameters such as fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to reactor dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties.

For example,

+/- = ( +/- ) (g +/- g) relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross section, ig, each with an uncertainty . The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least-squares evaluation of the dosimetry, the FERRET code (Reference 5) was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters along with associated uncertainties.

The application of the least-squares methodology requires the following input.

1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2. The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set.

3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the current application, the calculated neutron spectrum at each measurement location was obtained from the results of the previously described additional benchmarking neutron transport calculations. The spectrum at each sensor set location was input in an absolute sense (rather than simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements. The sensor reaction rates were derived from the measured specific activities of each sensor set and the operating history of the respective fuel cycles. The dosimetry reaction cross sections were obtained from the SNLRML dosimetry cross section library (Reference 6).

In addition to the magnitude of the calculated neutron spectra, the measured sensor set reaction 10 of 17

LR-N22-0012 LAR S22-01 rates, and the dosimeter set reaction cross sections, the least-squares procedure requires uncertainty estimates for each of these input parameters. The following provides a summary of the uncertainties associated with the least-squares evaluation of the dosimetry.

2.3 Additional Benchmarking Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, the irradiation history corrections, and the corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM International consensus standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input into the least-squares evaluation:

Reaction Uncertainty 63Cu (n,) Co 60 5%

46Ti (n,p) 46Sc 5%

54Fe (n,p) 54Mn 5%

58Ni (n,p) 58Co 5%

238U (n,f) 137Cs 10%

93Nb (n,n)93mNb 10%

237Np (n,f) 137Cs 10%

59Co (n,) 60Co 35%1 These uncertainties are given at the 1 level.

1 The cobalt content of older Co-Al foils used in EVND is not known for certain, but is believed to be between 0.438% and 0.562%. To account for this unknown, the uncertainty assigned in the least-squares evaluations (typically 5%) was increased by roughly (0.562 / 0.438 = 1.28) to 5% + 28% = 33%. Rounded to the nearest five, an uncertainty of 35% was input.

11 of 17

LR-N22-0012 LAR S22-01 2.4 Additional Benchmarking Dosimetry Cross Section Uncertainties As previously noted, the reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library. This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multi-group structure for use in least-squares adjustment applications. These cross sections were compiled from the ENDF/B-VI cross section evaluations and have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources. Detailed discussions of the contents of the SNLRML library along with the evaluation process for each of the sensors are provided in Reference 6.

For sensors included in the dosimetry sets, the following uncertainties in the fission spectrum-averaged cross sections are provided in the SNLRML documentation package:

Reaction Uncertainty 63Cu (n,) Co60 4.08-4.16%

46Ti (n,p) 46Sc 4.51-4.87%

54Fe (n,p) 54Mn 3.05-3.11%

58Ni (n,p) 58Co 4.49-4.56%

238U (n,f) 137Cs 0.54-0.64%

93Nb (n,n)93mNb 6.96-7.23%

237Np (n,f) 137Cs 10.32-10.97%

59Co (n,) 60Co 0.76-3.59%

These tabulated ranges provide an indication of the dosimetry cross section uncertainties associated with the sensor sets used in LWR irradiations.

2.5 Additional Benchmarking Calculated Neutron Spectrum Uncertainties While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks, and the dosimetry cross section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

= 2 +

Where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

= (1 ) +

Where:

( )2

=

2 2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range ( specifies the strength of the latter term). The value of is 1.0 when g = g and 0.0 otherwise.

The set of parameters defining the input covariance matrix for calculated spectra was as 12 of 17

LR-N22-0012 LAR S22-01 follows:

Flux Normalization Uncertainty (Rn) 15%

Flux Group Uncertainties (Rg, Rg)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short-Range Correlation ()

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range ()

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 These uncertainty assignments are consistent with an industry consensus uncertainty of 15 -

20% (1) for the fast neutron portion of the spectrum and provide for a reasonable increase in the uncertainty for neutrons in the intermediate and thermal energy ranges.

2.6 Additional Benchmarking Measurement-to-Calculation Comparison The comparison of the measurement results from each of the sensor set irradiations at RPV supports with corresponding analytical predictions at the measurement locations are presented in Table 6 and Table 7. These comparisons are provided on two levels. On the first level, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. This level of comparison is not impacted by the least-squares evaluations of the sensor sets. On the second level, calculated values of neutron exposure rates in terms of fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate are compared with the best-estimate exposure rates obtained from the least-squares evaluation.

In Table 6, comparisons of measurement-to-calculation (M/C) ratios are listed for the threshold sensors contained in the EVND dosimetry capsules irradiated at RPV supports that are approximately 8.5 feet above the core midplane. For the individual threshold foils, the average M/C ratio ranges from 0.49 to 1.33, with an overall average of 0.75 and an associated standard deviation of 28%. In this case, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Table 7, best-estimate-to-calculation (BE/C) ratios for fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate resulting from the least-squares evaluation of the dosimetry sets is provided for the EVND capsules irradiated at the RPV supports, which are approximately 8.5 feet above the core midplane. The BE/C ratio for the fast neutron (E > 1.0 MeV) fluence rate is 0.79 with an associated standard deviation of 19% and 0.88 with an associated standard deviation of 20% for the iron atom displacement rate (dpa/s). These BE/C ratios are within the +/-

20% uncertainty at 1- level required by Regulatory Guide (RG) 1.190 (Reference 7).

In summary, for the extended beltline region, the M/C data provided in Table 6 as well as the BE/C data provided in Table 7 suggest that the calculations are over predicting the neutron exposure, particularly at the high end of the energy spectrum. For instance, the bottom of the 90% neutron response for the copper, titanium, iron, and nickel dosimeters is 4.53 MeV, 13 of 17

LR-N22-0012 LAR S22-01 Enclosure 1 Attachment 3 3.70 MeV, 2.27 MeV, and 1.98 MeV, respectively. Neutrons with energies greater than these constitute a small fraction of the neutron (E > 1.0 MeV) fluence rate in the extended beltline region. The BE/C values in Table 7 account for the spectral coverage of the different sensors, and provide an estimate of the key damage parameters, fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate, that result from an uncertainty-weighted reconciliation of all of the measurements and calculations. The BE/C values in Table 7 suggest that the calculated damage parameters are moderately conservative relative to the best-estimate values.

Table 6: Measured-to-Calculated (M/C) Reaction Rates - Ex-Vessel Capsule Located in the Vicinity of the RPV Supports Capsule Capsule Capsule Capsule Capsule % Std.

Reaction E A K M R Average Dev.

63Cu (n,) 60Co 0.68 - 0.60 0.60 0.53 0.60 10%

46Ti (n,p) 46Sc 0.76 0.65 0.69 0.73 0.59 0.68 10%

54Fe (n,p) 54Mn 0.76 0.67 0.66 0.63 0.49 0.64 15%

58Ni (n,p) 58Co 0.78 0.68 0.68 0.74 0.54 0.68 13%

238U(Cd) (n,f) 1.09 0.92 - - - 1.01 12%

137 Cs 93Nb (n,n) 93mNb - - 1.33 1.02 0.58 0.98 39%

237Np(Cd) (n,f) 1.18 0.82 - - - 1.00 25%

137 Cs Average of M/C Results 0.75 28%

Table 7: Best-Estimate-to-Calculated (BE/C) Exposure Rates - Ex-Vessel Capsule Located in the Vicinity of the RPV Supports Neutron (E > 1.0 MeV) Fluence Iron Atom Displacement Rate Rate BE/C Capsule BE/C E 0.95 1.03 A 0.78 0.84 K 0.87 1.03 M 0.82 0.91 R 0.55 0.61 Average 0.79 0.88

% Std. Dev. 19% 20%

14 of 17

LR-N22-0012 LAR S22-01 3.0 Justification Conclusions The uncertainty associated with a fluence determination methodology is comprised of two major components: the results of an analytic uncertainty analysis and the results of benchmarking comparisons. An analytic uncertainty analysis assesses the level of confidence in key input parameters to a fluence calculation and quantifies the impact that plausible input parameter variations have on calculated fluence results. Benchmarking comparisons refer to comparisons of fluence calculations performed with a candidate methodology to alternate calculations or to measurements from a representative environment.

A comprehensive analytical uncertainty analysis for the Salem Unit 2 RPV extended beltline region demonstrates that the RAPTOR-M3G based calculations have a maximum of 63% uncertainty for the reported fast neutron fluence in the RPV extended beltline region of Salem Unit 2, which is associated with the lower shell to lower vessel head circumferential weld. Note that the uncertainty is increasing as the axial elevation of the RPV materials moving away from the active core. For elevations that are not as far from the top and bottom of the active core, the uncertainty of the fast neutron (E > 1.0 MeV) fluence determined using RAPTOR-M3G is much less than the maximum uncertainty stated above. Also, the uncertainty is much larger for the fast neutron fluence values reported at the outer radius of the RPV wall than those reported at the inner surface of the RPV wall. For example, the estimated uncertainty for the nozzle shell and nozzle shell to intermediate shell circumferential weld is approximately 30%. The Analytical Uncertainty for the Salem Unit 2 extended beltline materials are summarized in Table 8. This completes an important part of qualifying RAPTOR-M3G as the fluence determination methodology for RPV extended beltline region per RG 1.190 (Reference 7) regulatory position 1.4.1.

Table 8: Fast Neutron (E > 1.0 MeV) Fluence at 50 EFPY and Associated Analysis Uncertainty for the Extended Beltline Region Materials Axial Location to Core Midplane Unit 2 Analytical Material [cm] [n/cm2] Uncertainty Outlet Nozzle to Nozzle Shell 270.12 1.78E+16 35%

Welds - Lowest Extent Inlet Nozzle to Nozzle Shell Welds 265.04 2.26E+16 35%

- Lowest Extent Nozzle Shell 233.53 2.44E+17 30%

Nozzle Shell Longitudinal Welds 233.53 1.83E+17 30%

Nozzle Shell to Intermediate Shell 233.53 2.80E+17 30%

Circumferential Weld Lower Shell to Lower Vessel Head

-313.52 3.24E+15 63%

Circumferential Weld Additional benchmarking work described in Section 2.0 was performed at a 4-loop Westinghouse plant near the RPV supports that are approximately 8.5 feet above the core midplane. This work concluded that the RAPTOR-M3G fluence determination methodology has about a 30%

uncertainty in the fast neutron (E > 1.0 MeV) determination and the calculations typically overestimate the fast neutron (E > 1.0 MeV) fluence and iron atom displacement (dpa) at the extended beltline region. This completes the second important part of qualifying RAPTOR-M3G as the fluence determination methodology for RPV extended beltline region per RG 1.190 (Reference

7) regulatory position 1.4.2.

15 of 17

LR-N22-0012 LAR S22-01 The RPV extended beltline materials evaluated for Salem Unit 2 in Table 2 are all located within an axial distance of 8.86 feet above or below the core midplane. However, the lower shell to lower vessel head circumferential weld is approximately 1.42 feet (43.47 cm) further away. It is important to note that this material is not classified as an extended beltline material since the projected fluence values for Salem Unit 2 are well below 1E+17 n/cm2 and therefore not evaluated with respect to fracture toughness. This circumferential weld has a calculated fast neutron (E > 1.0 MeV) fluence of 3.24E+15 n/cm2 for Salem Unit 2 at 50 EFPY using RAPTOR-M3G, which is more than a factor of 30 lower than the prescribed threshold of 1E+17 n/cm2 for the definition of extended beltline region. Because the RAPTOR-M3G fluence determination methodology has a maximum of 63% uncertainty for the RPV extended beltline region this circumferential weld does not need to be included as extended beltline material that has to be evaluated for fracture toughness embrittlement effect.

From the discussion above, the methodology uncertainty for fast neutron (E > 1.0 MeV) fluence determination for these RPV extended beltline materials is also less than or equal to 63% for the fast neutron fluence values reported for Salem Unit 2. In review of downstream reactor vessel integrity analysis, significant margins in the fast neutron fluence on the extended beltline materials have been identified. For example, a review of RTPTS of the extended beltline materials determined the fluence would need to increase by 40 times in the Nozzle Shell (i.e., Upper Shell B4711-3) for it to become limiting. A review of 3/4T ART of the extended beltline materials determined the fluence would need to increase by 6 times in the Nozzle Shell (i.e., Upper Shell B4711-3) for it to becoming limiting. The estimated uncertainty using RAPTOR-M3G for the Nozzle Shell elevation is less than 30%. Therefore, it is not credible that any extended beltline region materials would ever become limiting for Salem Unit 2 at 50 EFPY. Finally, Section 3.4 of Reference 8 states that unless the fast neutron (E > 1.0 MeV) fluence for the nozzle material is greater than 4.28E+17 n/cm2, embrittlement need not be considered for nozzle forging evaluation and the nozzles will be non-limiting compared to the beltline with respect to the pressure-temperature limit curves.

Embrittlement was conservatively considered in the reactor vessel integrity analyses for nozzle materials, even if the fluence was below this threshold. The fast neutron (E > 1.0 MeV) fluence values reported for both the inlet and outlet nozzles in Table 2 are more than a factor of 18 lower than this fast neutron (E > 1.0 MeV) fluence threshold. As the evaluated net RAPTOR-M3G methodology uncertainty is approximately 35% or less for this elevation based on additional benchmarking and analytical uncertainty analysis, it is also not credible that the inlet and outlet nozzle fast neutron (E > 1.0 MeV) fluence at the EOLE will exceed 4.28E+17 n/cm2.

4.0 Application of WCAP-18124-NP-A, Revision 0 to the Extended Beltline Region - Conclusion Limitation and Condition #1 has been addressed in that the additional benchmarking at the RPV extended beltline region summarized herein, margin assessment documented in Reference 8, and the fluence analysis have provided additional justification supporting the use of the Reference 3 methodology for the extended beltline regions of the Salem Unit 2 RPV.

References:

1. Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, October 14, 2014.

(Agencywide Documents Access and Management System Accession Number ML14149A165) 16 of 17

LR-N22-0012 LAR S22-01

2. 10 CFR Appendix G to Part 50, Fracture Toughness Requirements.

3. Westinghouse Report, WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018.

4. RSICC Data Library Collection DLC-185, BUGLE-96 Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, March 1996. (Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.)

5. RSICC Computer Code Collection PSR-145 FERRET: Least-Squares Solution to Nuclear Data and Reactor Physics Problems, January 1980. (Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.)

6. RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross Section Compendium, July 1994. (Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.)

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

8. Pressurized Water Reactor Owners Group (PWROG) Report, PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020.

17 of 17

LR-N22-0012 LAR S22-01 Enclosure 2 Salem Unit 2 Pressure and Temperature Limits Report (PTLR)

PSEG Nuclear LLC Salem Generating Station Unit 2 Pressure and Temperature Limits Report (PTLR) for 50 Effective Full-Power Years (EFPY)

Revision 0 Prepared by: _Jk-3/4--'A-."'-=-~/2-.-.v-~--- Date: 7 - / 1... - 1,2..

'Sam\iel f~~

[Corporate Engineering Programs]

Reviewed by: A.A. fAkHA~ ~ 12.fl- A 1Tlfr-L iZfJ Date: 7--tl-- t..2 1

c.y'l,1 A ti Ali A. F akhar

[Manager, Corporate Engineering Programs]

Date: 7, 21 *;Zt).2 -l illiam J. Kopchick

[Sr. Director, Engineering Services]

Speer, Samuel E.

From: Fakhar, Ali A.

Sent: Thursday, July 21, 2022 7:29 AM To: Speer, Samuel E.

Subject:

Salem Unit 2 Pressure Temperature Limit Report (PTLR)

Sam, I have reviewed and approved the Salem Unit 2 Pressure-Temperature Limits Report.

Thanks, Ali 1

Salem Generating Station Unit 2 PTLR Revision 0 Page 2 of 40 Table of Contents Section Description Page 1.0 Purpose 3 2.0 Applicability 4 3.0 Operating Limits 5 4.0 Discussion 9 5.0 References 23 Figure 3-1 Salem Unit 2 Reactor Coolant System Heatup Limitations Applicable for Heatup Rates up to 100°F/HR for the Service Period up to 50 EFPY (with uncertainties for instrumentation errors) 7 Figure 3-2 Salem Unit 2 Reactor Coolant System Cooldown Limitations Applicable for Cooldown Rates up to 100°F/HR for the Service Period up to 50 EFPY (with instrumentation errors) 8 Table 4-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Salem Unit 2 Reactor Vessel Materials at 50 EFPY 12 Table 4-2 Salem Unit 2 Surveillance Capsule Data 15 Table 4-3 Calculation of Salem Unit 2 Chemistry Factors Using Surveillance Capsule Data 16 Table 4-4 Sister-Plant (Diablo Canyon Unit 2) Surveillance Program Results 17 Table 4-5 Calculation of Chemistry Factors Using Diablo Canyon Unit 2 Surveillance Capsule Data 17 Table 4-6 Limiting ART Values for Salem Unit 2 at 50 EFPY 20 Appendix A SALEM UNIT 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM 25 Appendix B TABULATED VALUES FOR SALEM UNIT 2 P-T CURVES 26 Appendix C SALEM UNIT 2 SURVEILLANCE DATA CREDIBILITY EVALUATION 33

Salem Generating Station Unit 2 PTLR Revision 0 Page 3 of 40 1.0 Purpose The purpose of the Salem Generating Station Unit 2 (Salem Unit 2) Pressure and Temperature (P-T) Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) P-T limits during Heat-up and Cooldown, inservice hydrostatic testing, and criticality;

2. RCS Heatup and Cooldown rates;

3. Reactor Pressure Vessel (RPV) head flange boltup temperature limits, and

4. Pressurizer Overpressure Protection System (POPS) setpoints and enable temperature.

This PTLR summarizes the various technical methodologies, equations, calculations, etc. that were necessary to produce the P-T Curves and POPS setpoints and enable temperature.

Throughout this PTLR, references will be made to the Westinghouse P-T limits analyses [1, 2]

that consist of the detailed explanations, figures, and tables.

This PTLR has been prepared in accordance with the requirements of Technical Specification 6.9.1.11.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 [10]. The revised PTLR shall be submitted to the NRC upon issuance.

Salem Generating Station Unit 2 PTLR Revision 0 Page 4 of 40 2.0 Applicability This report is applicable to the Salem Unit 2 RPV for up to 50 Effective Full-Power Years (EFPY).

The following Salem Unit 2 Technical Specifications (TS) are affected by the information contained in this report:

TS 3.1.2.3 Charging Pump - Shutdown TS 3.4.1.3 Reactor Coolant Loops - Hot Shutdown TS 3.4.1.4 Reactor Coolant Loops - Cold Shutdown TS 3.4.10.1 Reactor Coolant System Pressure/Temperature Limits TS 3.4.10.3 Reactor Coolant System Overpressure Protection Systems TS 3.5.3 ECCS Subsystems - Tavg < 350°F

Salem Generating Station Unit 2 PTLR Revision 0 Page 5 of 40 3.0 Operating Limits All limits are valid until 50 EFPY, which is projected to be beyond the expiration of the operating license for Salem Unit 2.

Pressurizer Overpressurization Overpressure Protection System (POPS) Enable Temperature 312°F*

The analyzed setpoint is 292°F, including instrument uncertainty, however, any changes to the setpoint of 312°F need to be evaluated under the Design Change Control Process.

Referenced in: TS 3.1.2.3, TS 3.4.10.1, TS 3.4.1.3, TS 3.4.1.4, TS 3.4.10.3, TS 3.5.3, and SR 4.5.3.2 RCS Pressure/Temperature (P/T) Limits Figure 3-1* RCS P-T Limits for Heatup Figure 3-2* RCS P-T Limits for Cooldown Referenced in: 3.4.10.1, SR 4.4.10.1.1, and SR 4.4.10.1.2

Figures include instrumentation uncertainty (18°F temperature margin and 61 psi pressure margin)

NOTE: Tables B-1, B-2, and B-3 of Appendix B contain a tabulated version of the curves.

Salem Generating Station Unit 2 PTLR Revision 0 Page 6 of 40 RCS Heatup/Cooldown Rate Limits Maximum heatup of 100°F in any one hour period.

Maximum cooldown of 100°F in any one hour period, and Maximum temperature change of less than or equal to 5°F in any one hour period during inservice hydrostatic testing operations above system design pressure.

Referenced in: TS 3.4.10.1, SR 4.4.10.1.1, and SR 4.4.10.1.2 Pressurizer Overpressurization Overpressure Protection System (POPS) Setpoint 375 psig*

Referenced in: TS 3.4.10.3

The analyzed setpoint is 434 psig, including instrument uncertainty, however, any changes to the setpoint of 375 psig need to be evaluated under the Design Change Control Process.

Minimum Boltup Temperature 62 °F*

Referenced in: PTLR Figures 3-1 and 3-2.

This temperature value accounts for 2°F temperature margin, Section 6.2 of [1].

Salem Generating Station Unit 2 PTLR Revision 0 Page 7 of 40 LIMITING MATERIAL AND ART VALUES AT 50 EFPY:

1/4T ART: 209°F (Lower Shell Longitudinal Weld Seams 3-442 A&C) 3/4T ART: 150°F (Lower Shell Longitudinal Weld Seams 3-442 A&C)

Operlim Version:5.2 Run:4802 Operlim.xls Version: 5.2 2500 Leak Test Limit 2250 2000 Unacceptable Acceptable Operation Operation 1750 1500 Calculated Pressure (PSIG) 1250 Critical Limit 60 Deg. F/Hr 1000 Critical Limit 100 Deg. F/Hr 750 Heatup Rate 60 Deg. F/Hr Criticality Limit based on 500 Heatup Rate inservice hydrostatic test 100 Deg. F/Hr temperature (289°F) for the service period up to 50 EFPY 250 Boltup Temperature = 62 F

~I (includes 2 F margin)

Lower limit for RCS pressure I I I is 0 psia without uncertainty 0

0 50 100 150 200 200 250 300 350 350 400 400 450 450 500 500 550 550 Moderator Temperature (Deg. F)

-250 Figure 3-1 Salem Unit 2 Reactor Coolant System Heatup Limitations Applicable for Heatup Rates up to 100°F/HR for the Service Period up to 50 EFPY (with uncertainties for instrumentation errors)

Salem Generating Station Unit 2 PTLR Revision 0 Page 8 of 40 LIMITING MATERIAL AND ART VALUES AT 50 EFPY:

1/4T ART: 209°F (Lower Shell Longitudinal Weld Seams 3-442 A&C) 3/4T ART: 150°F (Lower Shell Longitudinal Weld Seams 3-442 A&C) 2500 I Operlim Version:5.2 Run:4802 Operlim.xls Version: 5.2 I I

2250 2000 I j I

I-Unacceptable

~

Acceptable 1750 Operation I Operation I

1500 Calculated Pressure (PSIG) 1250 1

1000 J

Cooldown Rates 0 F/hr. (steady-state)

-20 F/hr.

-40 F/hr.

750 -60 F/hr.

-100 F/hr.

1 I

=E:t 1 ~ _y 500 I

250 Boltup Temperature = 62 F I

I, (includes 2 F margin)

J-+t l I I Lower limit for RCS pressure is 0 psia without uncertainty 0

0 50 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

-250 Figure 3-2 Salem Unit 2 Reactor Coolant System Cooldown Limitations Applicable for Cooldown Rates up to 100°F/HR for the Service Period up to 50 EFPY (with instrumentation errors)

Salem Generating Station Unit 2 PTLR Revision 0 Page 9 of 40 4.0 Discussion 4.1 Salem Unit 2 Surveillance Program Appendix A to this PTLR contains the Salem Unit 2 Surveillance Program.

4.2 General Overview of Development of P-T Limit Curves The analyses that support the generation of the Heatup and Cooldown Pressure-Temperature (P-T) limit curves presented in this PTLR are contained in References [1] and [2]. Reference [2] is a verification of the P-T limit curves originally developed in [1] for license renewal.

The P-T limit curves were generated through End of License Extension (EOLE), i.e., 60 years of operation, using the KIc methodology detailed in the 1998 through the 2000 Addenda Edition of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G [3]. The P-T limit curve generation methodology is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [4]. The heatup and cooldown P-T limit curves utilize the limiting Adjusted Reference Temperature (ART) values (RTNDT(U)) plus RTNDT plus margins for uncertainties) at the quarter thickness (1/4T) and three-quarter thickness (3/4T) locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface for Salem Unit 2. The ART values for 50 effective full-power years (EFPY) were calculated using Regulatory Guide 1.99, Revision 2 [5].

The Salem Unit 2 P-T limit curves were generated with instrumentation errors, and the reactor vessel flange requirements of 10 CFR 50, Appendix G [13].

4.3 Fluence Calculations To evaluate fast neutron exposure for the RPV, a three-dimensional discrete ordinates code, RAPTOR-M3G [7] was used, along with the BUGLE-96 cross-section library [14]. Radiation

Salem Generating Station Unit 2 PTLR Revision 0 Page 10 of 40 exposure parameters were established on a plant- and fuel-cycle-specific basis. Projections of future operation were based on the spatial power distributions and reactor operating conditions of Salem Unit 2 Cycles 20 and 23.

The neutron fluence was calculated as follows. Discrete ordinates (SN) transport analyses were performed to determine the neutron radiation environment within the RPV. In these analyses, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis. The dosimetry analysis documented in Appendix C of [2] shows that the +/-20% (1) acceptance criteria specified in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [6], is met, based on the measurement-to-calculation (M/C) comparison results for the in-vessel surveillance capsules withdrawn and analyzed to-date. These validated calculations form the basis for providing projections of the neutron exposure of the RPV through EOLE.

All of the calculations were based on nuclear cross-section data derived from the Evaluated Nuclear Data File (ENDF) database (specifically, ENDF/B-VI). Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET [7]. The neutron transport evaluation methodology is based on the guidance of Regulatory Guide 1.190.

The uncertainty associated with the calculated neutron exposure of the Salem Unit 2 RPV is based on the recommended approach provided in Regulatory Guide 1.190. The qualification of the methodology used in the plant-specific neutron exposure evaluation is carried out in the following four stages:

Comparisons of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator (NUREG/CR-6454, "Pool Critical Assembly Pressure Vessel Facility Benchmark" [8]) at the Oak Ridge National Laboratory (ORNL) and the VENUS-1 experiment.

Salem Generating Station Unit 2 PTLR Revision 0 Page 11 of 40 Comparison of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment (NUREG/CR-6453, "H.B. Robinson 2 Pressure Vessel Benchmark" [9]).

An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments (WCAP-18124-NP-A, "Fluence Determination with RAPTOR-M3G and FERRET" [7]).

Comparison of calculations with all available dosimetry results from the RPV measurement programs carried out at Salem Unit 2 (Appendix C of [2]).

Table 4-1 (re-produced from Table 6-1 of [2]) summarizes the fluence projections at 50 EFPY for the Vessel Surface, 1/4T, and 3/4T locations for the Salem Unit 2 RPV. These fluence values were used to calculate the end of life extension (EOLE) ART values for the Salem Unit 2 RPV.

Salem Generating Station Unit 2 PTLR Revision 0 Page 12 of 40 Table 4-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Salem Unit 2 Reactor Vessel Materials at 50 EFPY Surface 1/4T 3/4T fluence(a) fluence(b) fluence(b)

Material Surface FF(c) 1/4T FF(c) 3/4T FF(c)

(x 1019 n/cm2, (x 1019 n/cm2, (x 1019 n/cm2, E> 1.0 MeV) E> 1.0 MeV) E> 1.0 MeV)

Upper Shell Plates 0.0244 0.194 0.0145 0.141 0.00517 0.069 Intermediate Shell Plates 2.05 1.196 1.22 1.056 0.434 0.768 Lower Shell Plates 2.01 1.190 1.20 1.050 0.426 0.763 Intermediate to Upper Shell Circumferential Weld Seam 0.028 0.211 0.0167 0.154 0.00593 0.076 8-442 Intermediate to Lower Shell Circumferential Weld Seam 1.98 1.186 1.18 1.046 0.419 0.759 9-442 Upper Shell Longitudinal Weld 0.0183 0.163 0.0109 0.116 0.00387 0.055 Seams 1-442A & C Upper Shell Longitudinal Weld 0.0101 0.110 0.0060 0.077 0.00214 0.035 Seam 1-442B Intermediate Shell Longitudinal 1.47 1.107 0.876 0.963 0.311 0.680 Weld Seams 2-442 B & C Intermediate Shell Longitudinal 0.741 0.916 0.442 0.773 0.157 0.513 Weld Seam 2-442A Lower Shell Longitudinal Weld 1.45 1.103 0.864 0.959 0.307 0.676 Seams 3-442 A & C Lower Shell Longitudinal Weld 0.738 0.915 0.440 0.772 0.156 0.512 Seams 3-442B Note(s):

(a) Fluence values are documented in Table 2-2 of [2].

(b) The 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (8.625 inches) and equation f = fsurf

e-0.24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).

(c) FF = fluence factor = f(0.28 - 0.10*log (f)).

Salem Generating Station Unit 2 PTLR Revision 0 Page 13 of 40 4.4 Fracture Toughness Properties P-T limit curve development requirements are specified in 10 CFR 50, Appendix G [13], which also defines the beltline region of the RPV. Any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period should be considered in the development of P-T limit curves, where these additional materials that exceed this fluence threshold are referred to as the extended beltline materials and are evaluated to ensure that the applicable neutron embrittlement effects are considered.

For Salem Unit 2, the extended beltline materials include upper shell plates, upper shell longitudinal welds, and the upper to intermediate shell girth weld. However, the fluence for both inlet/outlet nozzle to upper shell welds are less than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at 50 EFPY, therefore, the nozzle forgings and associated welds to the upper shell do not need to be considered in the extended beltline.

Note: for RPV welds, the terms girth and circumferential are used interchangeably and for the purposes of this PTLR and its supporting documentation, these welds shall be referred to as circumferential welds. Similarly, the terms axial and longitudinal are used interchangeably; herein, these welds are referred to as longitudinal welds.

Although the RPV nozzles are not a part of the extended beltline, per NRC RIS 2014-11 [11], the nozzle materials must be evaluated for their potential effect on P-T limit curves due to the higher stresses in the nozzle corner region. The effects of these higher stresses are addressed in [2],

which determines that the Salem Unit 2 beltline P-T limit curves - bound

- - - the inlet and outlet nozzle P-T limit curves.

Table 3-1 of [2] contains the summary of the best-estimate copper and nickel contents in units of weight percent (wt. %), as well as initial RTNDT values, for the Salem Unit 2 reactor vessel beltline and extended beltline materials. Table 3-2 of [2] contains the summary of the best-

Salem Generating Station Unit 2 PTLR Revision 0 Page 14 of 40 estimate copper and nickel contents in units of weight percent (wt. %), as well as initial RTNDT values, for the Salem Unit 2 reactor vessel inlet and outlet nozzle materials. Table 3-1 of [1] lists the initial RTNDT values for the Replacement Reactor Vessel Closure Head (RRVCH) and RPV flange materials.

4.5 Use of Surveillance Data Position 2.1 of RG 1.99, Rev. 2 [5] requires data from the plant-specific surveillance program to calculate chemistry factors. Table 4-1 of [2] lists the capsule surveillance data for Salem Unit 2 to date, and Table 4-2 of [2] lists data from a surveillance program at a sister plant, Diablo Canyon Unit 2, which includes a Salem Unit 2 RPV extended beltline material that should also be considered when calculating Position 2.1 chemistry factors. Appendix C of this PTLR reproduces the credibility re-evaluation of the surveillance data as contained in Appendix A of

[2].

4.6 Chemistry Factors The chemistry factors (CFs) were calculated using Regulatory Guide 1.99, Revision 2 [5],

Positions 1.1 and 2.1. Since Salem Unit 2 has a plant-specific surveillance program, Position 2.1 chemistry factor calculations were performed using the method described in Regulatory Guide 1.99, Revision 2 [5]. Table 4-2 of this PTLR (re-produced from Table 4-1 of [2]) summarizes the surveillance data available for the Salem Unit 2 plate and weld materials that were used in the calculation of the Position 2.1 chemistry factor values. The Position 2.1 chemistry factor calculations are presented in Table 4-3 of this PTLR (re-produced from Table 5-1 of [2]) for the Salem Unit 2 surveillance materials.

Salem Unit 2 considered surveillance data from a sister plant, Diablo Canyon Unit 2, for weld heat # 21935/12008. Adjustment of the RTNDT values were required per Regulatory Guide 1.99, Revision 2 [5] due to chemistry and irradiation temperature differences between the surveillance welds and Salem Unit 2 RPV welds. Position 2.1 chemistry factor calculations

Salem Generating Station Unit 2 PTLR Revision 0 Page 15 of 40 using the Diablo Canyon Unit 2 surveillance data for weld heat # 21935/12008 were performed using the method described in Regulatory Guide 1.99, Revision 2 [5]. Table 4-4 of this PTLR (re-produced from Table 4-2 of [2]) summarizes the surveillance data available for the Diablo Canyon Unit 2 weld materials that were used in the calculation of the Position 2.1 chemistry factor values. The Position 2.1 chemistry factor calculations are presented in Table 4-5 of this PTLR (re-produced from Table 5-2 of [2]) for the Diablo Canyon Unit 2 surveillance materials.

Table 4-2 Salem Unit 2 Surveillance Capsule Data(a)

Fluence(b)

Measured Irradiation (n/cm2, Material Capsule RTNDT Temperature E > 1.0

(°F) (°F)

MeV)

T 0.273E+19 61.66 545 Intermediate Shell Plate B4712-2 U 0.581E+19 66.54 543 (Longitudinal) X 1.13E+19 93.82 541 Y 1.83E+19 105.69 540 T 0.273E+19 74.83 545 Intermediate Shell Plate B4712-2 U 0.581E+19 98.26 543 (Transverse) X 1.13E+19 125.15 541 Y 1.83E+19 129.33 540 T 0.273E+19 153.17 545 Salem Unit 2 U 0.581E+19 185.94 543 Surveillance Weld Material(c)

(Heat # 13253) X 1.13E+19 195.43 541 Y 1.83E+19 200.90 540 Note(s):

(a) Information extracted from WCAP-15692 {refer to Ref. 15 of [1]}, unless otherwise noted.

(b) The fluence values were taken from Table 2-4 of [2].

(c) Note that the Salem Unit 2 surveillance weld metal Heat # 13253 is only representative of the beltline welds, not identical.

Therefore, the surveillance weld data was not used in the calculations documented in [1] and [2].

Salem Generating Station Unit 2 PTLR Revision 0 Page 16 of 40 Table 4-3 Calculation of Salem Unit 2 Chemistry Factors Using Surveillance Capsule Data(a)

Capsule Measured Material Capsule Fluence FF(b) FF*RTNDT FF2 (x 1019 n/cm2, RTNDT (°F)

(°F)

E > 1.0 MeV)

Intermediate T 0.273 0.646 61.66 39.84 0.418 Shell Plate U 0.581 0.848 66.54 56.43 0.719 B4712-2 X 1.13 1.034 93.82 97.02 1.069 (Longitudinal) Y 1.83 1.199 105.69 123.21 1.359 Intermediate T 0.273 0.646 74.83 48.35 0.418 Shell Plate U 0.581 0.848 98.26 83.33 0.719 B4712-2 X 1.13 1.034 125.15 129.42 1.069 (Transverse) Y 1.83 1.199 129.33 150.76 1.359 SUM: 728.36 7.130 2

CFB4712-2 = (FF

RTNDT) ÷ (FF ) = (728.36) ÷ (7.130) = 102.2°F T 0.273 0.646 153.17 98.97 0.418 Salem Unit 2 U 0.581 0.848 185.94 157.68 0.719 Surveillance X 1.13 1.034 195.43 202.10 1.069 Weld Material Y 1.83 1.199 200.90 234.20 1.359 (Heat # 13253) SUM: 692.95 3.565 CFSurv. Weld = (FF

RTNDT) ÷ (FF2) = (692.95) ÷ (3.565) = 194.4°F Note(s):

(a) Data taken from Table 4-1 of [2], unless otherwise noted.

(b) FF = fluence factor = f(0.28 - 0.101*log (f)).

Salem Generating Station Unit 2 PTLR Revision 0 Page 17 of 40 Table 4-4 Sister-Plant (Diablo Canyon Unit 2) Surveillance Program Results Withdrawal Capsule Fluence Measured Irradiation Cycle EFPY (x 1019 Material Capsule RTNDT Temperature n/cm2,

(°F) (°F)

E > 1.0 MeV)

Diablo Canyon Unit 2 U 1 1.02 0.330 173.0 545 Surveillance Weld X 3 3.16 0.906 203.2 545 Material(a) Y 6 7.08 1.53 211.4 545 (Heat # 21935/12008) V 9 11.49 2.38 224.5 545 Note(s):

(a) The Diablo Canyon Unit 2 capsule data were taken from WCAP-17315-NP (Ref. 15 of [2]). EFPY values are found in WCAP-15423 (Ref. 12 of [2]).

Table 4-5 Calculation of Chemistry Factors Using Diablo Canyon Unit 2 Surveillance Capsule Data(a)

Capsule Measured Material Capsule Fluence FF(b) FF*RTNDT FF2 (x 1019 n/cm2, RTNDT(b) (°F)

(°F)

E > 1.0 MeV) 176.0 122.32 0.483 U 0.330 0.695 (173.0)

Diablo Canyon 206.2 200.49 0.945 X 0.906 0.972 Unit 2 (203.2)

Surveillance 214.4 239.62 1.249 Y 1.53 1.118 Weld Material (211.4)

(Heat # 227.5 280.70 1.522 V 2.38 1.234 21935/12008) (224.5)

SUM: 843.14 4.200 CFSurv. Weld = (FF

RTNDT) ÷ (FF2) = (843.14) ÷ (4.200) = 200.7°F Note(s):

(a) Fluence and RTNDT taken from Table 4-2 of [2].

(b) FF = fluence factor = f(0.28 - 0.101*log (f))

(c) The surveillance weld RTNDT values have been adjusted for irradiation temperature as follows:

Adjusted RTNDT = RTNDT, Measured + Temp. Adjustment The temperature adjustments are based on a Salem Unit 2 reactor vessel temperature of 542°F. The Diablo Canyon Unit 2 capsule irradiation temperatures are provided in Table 4-2 of [2]. The measured (unadjusted) RTNDT values are shown in parenthesis. The temperature adjustment is therefore 3°F (545°F - 542°F). Note, CF ratio procedure (CFSalem 2 vessel weld / CFDiablo Canyon 2 surv. weld) results in a ratio of less than 1.0; therefore, the chemistry adjustment can conservatively be neglected.

Salem Generating Station Unit 2 PTLR Revision 0 Page 18 of 40 4.7 Criteria for Allowable Pressure-Temperature Relationships The approach to the development of the Salem Unit 2 P-T limit curves, including their equations and figures from the ASME Code,Section XI, Appendix G is contained in [1]. The P-T limit curve methodology is the same as that described in WCAP-14040-A [4]. The RPV metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A.

The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* KIm + KIt < KIc

where, KIm = stress intensity factor caused by membrane (pressure) stress KIt= stress intensity factor caused by the thermal gradients KIc= reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT C= 2.0 for Level A and Level B service limits C= 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, KIc is determined by using the metal temperature at the tip of a postulated flaw (1/4T) and the RTNDT, in the reference fracture toughness curve equation, Equation 2.5-2 from [4]. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, for the reference flaw were computed, the pressure stress intensity factors, KIm, were then obtained, and from these, the allowable pressures were calculated.

Salem Generating Station Unit 2 PTLR Revision 0 Page 19 of 40 4.8 Closure Head / Vessel Flange Requirements Regarding RPV closure head flange and RPV flange regions, 10 CFR Part 50, Appendix G [13]

states that the metal temperature of the closure head regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which was calculated to be 621 psig [1]. The RPV closure head and RPV flange initial RTNDT values (shown as RTNDT(U)) are provided in [1]. The limiting unirradiated RTNDT of 12°F (rounded up to 15°F ) is associated with the vessel flange of the Salem Unit 2 RPV, so the minimum allowable temperature of this region is 153°F for normal operation at pressures greater than 560 psig with margins for instrument uncertainties.

4.9 Boltup Temperature Requirements Boltup temperature requirements are established in Appendix G to 10 CFR 50 [13]. Per [4], the minimum boltup temperature should be 60°F or the limiting unirradiated RTNDT of the RPV closure flange region, whichever is higher. The minimum boltup temperature for the Salem Unit 2 RPV is 60°F without margins for instrument uncertainties and 62°F with margins for instrument uncertainties.

4.10 Calculation of ART The formulas and methodology for calculating the end of life extension (EOLE) ART values, which are based on fluence calculated at 50 EFPY, are provided in [2]. Salem Unit 2 had generated P-T Curves as part of license renewal [1], but were not submitted for approval at the time. In order to support the implementation of the 50 EFPY P-T limit curves in Figure 3-1 and 3-2 of this PTLR, the applicability of the P-T limit curves was evaluated by comparing the 50 EFPY limiting 1/4T and 3/4T ART values contained in Tables 6-2 and 6-3 of [2] with those used developed in [1].

Salem Generating Station Unit 2 PTLR Revision 0 Page 20 of 40 Table 4-6 contains the limiting ART values for the Salem Unit 2 RPV at 50 EFPY (partially reproduced from Table 7-1 of [2]).

Table 4-6 Limiting ART Values for Salem Unit 2 at 50 EFPY Verification Analysis(a) P-T Limit Curves Analysis(b)

Limiting ART Limiting ART Limiting Material Limiting Material Values Value)

Intermediate Shell Longitudinal Weld Seams 2- Lower Shell Longitudinal 1/4T Location 191.6(c) 209 Weld Seam 3-442 A & C 442 B & C (Position 1.1)

(Position 1.1)

Intermediate Shell Longitudinal Weld Seams 2- Lower Shell Longitudinal 3/4T Location 138.1(c) 442 B & C 150 Weld Seam 3-442 A & C (Position 1.1) (Position 1.1)

Note(s):

(a) Values are the limiting values from Tables 6-2 and 6-3 of [2].

(b) Values are the limiting values as listed in Table 6-1 of [1]. The P-T Curves developed in [1] are based on more conservative ART values computed in [2], therefore, are the governing curves for this PTLR.

(c) Note, the ART value calculated for Lower Shell Longitudinal Weld Seams 3-442 A & C have higher ART values at both the 1/4T and 3/4T locations when a Position 1.1 CF is used [1]. Because credible surveillance data exists from the Diablo Canyon Unit 2 (sister plant) surveillance program, the lower 1/4T and 3/4T ART values, calculated with a Position 2.1 CF, as documented in [2], supersede those values. Therefore, the Limiting ART values to use in [2] for the P-T Curve applicability are 191.6°F and 138.1°F for 1/4T and 3/4T locations, respectively.

The P-T limit curves developed in [1] remain valid since they are based on the bounding ART values listed above in Table 4-6.

4.11 Pressurized Thermal Shock (PTS) 10 CFR 50.61 establishes screening criteria on pressurized water reactor (PWR) vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTS. RTPTS screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld

Salem Generating Station Unit 2 PTLR Revision 0 Page 21 of 40 seams for plant operation to the end of plant license. The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions make the procedure for calculating the RTPTS values consistent with the methods given in Regulatory Guide 1.99, Revision 2.

These accepted methods were used with the surface fluence values to calculate the following RTPTS values for the Salem Unit 2 RPV materials. The results of the RTPTS calculations are presented in Table B-1 of [2] for Salem Unit 2.

All of the reactor vessel materials for Salem Unit 2 are below the RTPTS screening criteria values of 270°F for plates, forgings, and longitudinal welds, and 300°F for circumferentially oriented welds (per 10 CFR 50.61) at 50 EFPY.

4.12 Heatup and Cooldown P-T Limit Curves For P-T limit curve development, the limiting ART values from Table 4-4 were used, i.e., 209°F and 150°F for the limiting 1/4T and 3/4T locations, respectively.

The Salem Unit 2 P-T limit curves are presented in Section 3 of this PTLR, which incorporate all necessary limits for ensuring prevention of non-ductile failure for the Salem Unit 2 RPV with the flange requirements and with instrumentation uncertainties. The data points used to construct the P-T limit curves are shown in Tables B-1, B-2, and B-3 of Appendix B to this PTLR. Vacuum refill limits for the RCS are depicted on the P-T Curves by showing a minimum pressure of 0 psia.

Salem Generating Station Unit 2 PTLR Revision 0 Page 22 of 40 4.13 LTOP Enable Temperature ASME Code Case N-641 [15] includes a provision for calculating P-T relationships and low temperature overpressure protection (LTOP) system effective temperatures, Te, and allowable pressures. The enable temperature presented in [1] is based on the same ART values as the P-T limit curves. Since the P-T limit curves developed in [1] remain valid, the enable temperature reported in [1] also remains valid. Thus, the minimum required enable temperature without margins for instrument uncertainty is a coolant temperature equal to 274°F through EOLE. With margins for instrument uncertainty (+18°F), the minimum required enable temperature is a coolant temperature equal to 292°F. This analyzed enable temperature is lower than the current 32 EPFY setpoint of 312°F (more conservative), which will remain as the enable temperature setpoint as stated in Section 3.0 of this PTLR. Salem Unit 2 may revise the enable temperature setpoint to a value down to 292°F following an approved design change.

4.14 Pressurizer Overpressurization Protection System (POPS) Analysis The LTOPS is also known as the POPS at Salem Unit 2 and consists of the two Power Operated Relief Valves (PORVs) with reduced relief settings. These setpoints were selected following the NRC-approved methodology in WCAP-14040-A, Rev. 4 [4] such that the peak pressure during the design basis Mass Injection (MI) and Heat Injection (HI) transients will not exceed the isothermal 10CFR50 Appendix G [13] P-T limits.

The analyses that support the generation of the POPS PORV setpoint presented in this PTLR are contained in References [16] and [17]. Reference [16] is a verification of the POPS setpoint originally developed in [17] for license renewal.

The final maximum allowable PORV setpoint is determined such that it bounds both the MI and HI transient maximum allowable PORV setpoints.

Salem Generating Station Unit 2 PTLR Revision 0 Page 23 of 40 The analyzed Salem Unit 2 POPS PORV setting is 434 psig [16], which is higher than the current 32 EPFY setpoint of 375 psig (more conservative) as stated in Section 3.0 of this PTLR.

Salem Unit 2 may revise the POPS PORV setpoint to a value up to 434 psig following an approved design change.

5.0 References

1. PSEG VTD 901764 (001), Westinghouse WCAP-16982-NP, Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity (Non-Proprietary Class 3), Rev. 3, December 2020

2. PSEG VTD 904297 (001), Westinghouse WCAP-18571-NP, Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation (Non-Proprietary Class 3), Rev. 2, July 2022

3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."

4. Westinghouse Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004

5. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

6. Regulatory Guide 1.190, Revision 0, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, March 2001.

7. Westinghouse Report WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018.

Salem Generating Station Unit 2 PTLR Revision 0 Page 24 of 40

8. ORNL Report ORNL/TM-13205, Pool Critical Assembly Pressure Vessel Facility Benchmark, (NUREG/CR-6454), July 1997.

9. ORNL Report ORNL/TM-13204, H. B. Robinson-2 Pressure Vessel Benchmark, (NUREG/CR 6453), October 1997 (published February 1998).

10. 10CFR50.59, Changes, tests, and experiments.

11. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S. Nuclear Regulatory Commission, October 2014.

12. ASME Boiler and Pressure Vessel Code Section III.

13. Appendix G to Part 50Fracture Toughness Requirements.

14. RSICC Data Library Collection DLC-185, BUGLE-96: Coupled 47 Neutron, 20 GammC-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, July 1999.

15. ASME Code Case N-641.

16. PSEG VTD 904298 (001), Westinghouse Letter Report LTR-SCS-20-52-P, Rev. 1, Salem Unit 2 Low Temperature Overpressure Protection System (LTOPS) / Pressurizer Overpressure Protection System (POPS) Analysis, July 26, 2021.

17. PSEG VTD 904298 (003), Westinghouse Letter PNG-LTP-TR-AA-000001-P, Rev. 1, Transmittal of PSE-09 Low Temperature Overpressure Protection System (LTOPS) Setpoint Analysis for Salem Unit 1 and 2 to Support the Salem Unit 2 LAR.

Salem Generating Station Unit 2 PTLR Revision 0 Page 25 of 40 Appendix A SALEM UNIT 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM The Salem Unit 2 Surveillance Program is described in UFSAR Section 5.2.4.4.

In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [A-1], four (4) surveillance capsules have been removed from the Salem Unit 2 reactor vessel in accordance with the following schedule.

Capsule T Removed in 1983 WCAP-10492, Analysis of Capsule T from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, March 1984 Capsule U Removed in 1986 WCAP-11554, Analysis of Capsule U from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, September 1987 Capsule X Removed in 1991 WCAP-13366, Analysis of Capsule X from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, June 1992 Capsule Y Removed in 2000 WCAP-15692, Analysis of Capsule Y from the Public Service Electric and Gas Co. Salem Unit 2 Reactor Vessel Radiation Surveillance Program, August 2001 Capsule S Scheduled to be removed at 32 EFPY, tentatively May 2023 Capsule V Standby Capsule W Standby Capsule Z Standby APPENDIX A

REFERENCES:

A-1 U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, October 2, 2020.

Salem Generating Station Unit 2 PTLR Revision 0 Page 26 of 40 Appendix B TABULATED VALUES FOR SALEM UNIT 2 P-T CURVES

Salem Generating Station Unit 2 PTLR Revision 0 Page 27 of 40 Table B-1 (Re-produced from Table 6-2 of [1])

50 EFPY Heatup Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature]

60°F/hr 60°F/hr 100°F/hr 100°F/hr Heatup Criticality Heatup Criticality T P T P (psig) T P T P

(°F) (psig) (°F) (°F) (psig) (°F) (psig) 62 Note (a) 289 Note (a) 62 Note (a) 289 Note (a) 62 555 289 555 62 510 289 510 83 555 289 555 83 510 289 511 88 555 289 556 88 510 289 511 93 555 289 558 93 510 289 513 98 555 289 560 98 510 289 514 103 555 289 560 103 510 289 516 108 555 289 560 108 510 289 518 113 555 289 560 113 510 289 521 118 556 289 560 118 510 289 523 123 560 289 560 123 510 289 528 128 560 289 560 128 510 289 531 133 560 289 560 133 510 289 536 138 560 289 560 138 511 289 541 143 560 289 560 143 513 289 546 148 560 289 560 148 516 289 552 153 560 289 613 153 521 289 557 153 560 289 627 153 521 289 560 153 613 289 644 153 521 289 560 158 627 289 663 158 528 289 560 163 644 289 684 163 536 289 560 168 662 289 707 168 546 289 570 173 675 289 733 173 557 289 585

Salem Generating Station Unit 2 PTLR Revision 0 Page 28 of 40 Table B-1 (Re-produced from Table 6-2 of [1])

50 EFPY Heatup Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology

[With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature]

60°F/hr 60°F/hr 100°F/hr 100°F/hr Heatup Criticality Heatup Criticality T P T P (psig) T P T P

(°F) (psig) (°F) (°F) (psig) (°F) (psig) 178 688 289 762 178 570 289 603 183 703 289 794 183 585 289 622 188 720 289 829 188 603 289 644 193 738 289 868 193 622 289 669 198 758 289 906 198 644 289 696 203 781 289 927 203 669 289 727 208 805 289 950 208 696 289 761 213 833 289 976 213 727 289 798 218 863 289 1003 218 761 289 840 223 896 289 1034 223 798 289 886 228 933 289 1068 228 840 289 938 233 973 289 1105 233 886 289 994 238 1018 289 1132 238 938 289 1067 243 1068 293 1184 243 994 293 1126 248 1123 298 1241 248 1057 298 1203 253 1184 303 1296 253 1126 303 1287 258 1241 308 1356 258 1203 308 1354 263 1296 313 1423 263 1287 313 1409 268 1356 318 1496 268 1354 318 1470 273 1423 323 1577 273 1409 323 1537 278 1496 328 1667 278 1470 328 1611 283 1577 333 1766 283 1537 333 1692

Salem Generating Station Unit 2 PTLR Revision 0 Page 29 of 40 Table B-1 (Re-produced from Table 6-2 of [1])

50 EFPY Heatup Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology

[With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature]

60°F/hr 60°F/hr 100°F/hr 100°F/hr Heatup Criticality Heatup Criticality T P T P T P T P

(°F) (psig) (°F) (psig) (°F) (psig) (°F) (psig) 288 1667 338 1875 288 1611 338 1782 293 1766 343 1995 293 1692 343 1880 298 1875 348 2128 298 1782 348 1989 303 1995 353 2274 303 1880 353 2109 308 2128 308 1989 358 2241 313 2274 313 2109 363 2386 318 2241 323 2386 Note:

(a) The lower limit for reactor coolant system pressure is 0 psia without uncertainty.

Salem Generating Station Unit 2 PTLR Revision 0 Page 30 of 40 Table B-2 (Re-produced from Table 6-3 of [1])

50 EFPY Cooldown Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature]

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

T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 62 Note (a) 62 Note (a) 62 Note (a) 62 Note (a) 62 Note (a) 62 560 62 515 62 464 62 411 62 302 83 560 83 517 83 466 83 413 83 304 88 560 88 519 88 468 88 415 88 306 93 560 93 522 93 470 93 418 93 309 98 560 98 525 98 473 98 421 98 313 103 560 103 528 103 476 103 424 103 316 108 560 108 531 108 480 108 428 108 321 113 560 113 535 113 484 113 432 113 326 118 560 118 539 118 489 118 437 118 331 123 560 123 544 123 494 123 442 123 338 128 560 128 549 128 499 128 449 128 345 133 560 133 555 133 506 133 455 133 353 138 560 138 560 138 513 138 463 138 362 143 560 143 560 143 520 143 471 143 372 148 560 148 560 148 529 148 481 148 383 153 560 153 560 153 539 153 491 153 396 153 560 153 560 158 549 158 503 158 410 153 632 153 586 163 561 163 516 163 426 158 641 158 595 168 574 168 530 168 444 163 651 163 606 173 589 173 546 173 463 168 662 168 618 178 605 178 564 178 485 173 675 173 632 183 623 183 584 183 510 178 688 178 646 188 643 188 606 188 537 183 703 183 663 193 665 193 630 193 567

Salem Generating Station Unit 2 PTLR Revision 0 Page 31 of 40 Table B-2 (Re-produced from Table 6-3 of [1])

50 EFPY Cooldown Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature]

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

T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 188 720 188 681 198 690 198 658 198 601 193 738 193 701 203 717 203 688 203 638 198 758 198 723 208 747 208 721 208 680 203 781 203 748 213 780 213 758 213 726 208 805 208 775 218 817 218 799 218 777 213 833 213 805 223 858 223 844 223 834 218 863 218 838 228 903 228 895 228 895 223 896 223 875 233 953 233 950 233 950 228 933 228 916 238 1008 238 1008 238 1008 233 973 233 961 243 1066 243 1066 243 1066 238 1018 238 1011 248 1123 248 1123 248 1123 243 1068 243 1066 253 1184 253 1184 253 1184 248 1123 248 1123 258 1251 258 1251 258 1251 253 1184 253 1184 263 1325 263 1325 263 1325 258 1251 258 1251 268 1407 268 1407 268 1407 263 1325 263 1325 273 1497 273 1497 273 1497 268 1407 268 1407 278 1597 278 1597 278 1597 273 1497 273 1497 283 1708 283 1708 283 1708 278 1597 278 1597 288 1830 288 1830 288 1830 283 1708 283 1708 293 1965 293 1965 293 1965 288 1830 288 1830 298 2115 298 2115 298 2115 293 1965 293 1965 303 2280 303 2280 303 2280 298 2115 298 2115 303 2280 303 2280 Note:

(a) The lower limit for reactor coolant system pressure is 0 psia without uncertainty.

Salem Generating Station Unit 2 PTLR Revision 0 Page 32 of 40 Table B-3 (Re-produced from Table 6-2 of [1])

Salem Unit 2 50 EFPY Inservice Leak Test Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (with KIc, and Margins for Temperature (18°F) and Pressure (61 psi) Instrumentation Errors)

Temperature Pressure

(°F) (psig) 272 2000 289 2485

Salem Generating Station Unit 2 PTLR Revision 0 Page 33 of 40 Appendix C SALEM UNIT 2 SURVEILLANCE DATA CREDIBILITY EVALUATION

Salem Generating Station Unit 2 PTLR Revision 0 Page 34 of 40 Regulatory Guide 1.99, Revision 2 [Ref. C-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describe the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data.

The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date, there have been four surveillance capsules removed and tested from the Salem Unit 2 reactor vessel. In addition, surveillance data from a sister plant is utilized herein to perform neutron radiation embrittlement calculations.

Table C-1 reviews the five criteria of Regulatory Guide 1.99, Revision 2. The following subsections evaluate each of these five criteria for Salem Unit 2 in order to determine the credibility of the surveillance data for use in neutron radiation embrittlement calculations.

Table C-1 Identification of Credibility Criteria Affected by Fluence Change Criterion Description No.

Materials in the capsules should be those judged most likely to be controlling with regard to radiation 1

embrittlement.

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions 2 should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

When there are two or more sets of surveillance data from one reactor, the scatter of RT NDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of 3

magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. C-4].

The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall 4

temperature at the cladding/base metal interface within +/- 25°F.

The surveillance data for the correlation monitor material in the capsule should fall within the scatter 5

band of the database for that material.

Salem Generating Station Unit 2 PTLR Revision 0 Page 35 of 40 Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR 50, Fracture Toughness Requirements, [Ref. C-2] as follows:

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

The Salem Unit 2 reactor vessel consists of the following beltline region materials, which likely would have been considered at the time the surveillance program was designed and licensed:

Intermediate Shell Plates B4712-1, B4712-2, and B4712-3 Lower Shell Plates B4713-1, B4713-2, and B4713-3 Intermediate to Lower Shell Circumferential Weld Seam 9-442 (Heat # 90099, Linde 0091 flux, Lot # 3977)

Intermediate Shell Longitudinal Weld Seams 2-442 A, 2-442 B, and 2-442 C (Heat # 13253/20291, Linde 1092 flux, Lot # 3833)

Lower Shell Longitudinal Weld Seams 3-442 A, 3-442 B, and 3-442 C (Heat # 21935/12008, Linde 1092 flux, Lot # 3889)

Intermediate Shell Plate B4712-2 has the highest initial RTNDT of the beltline plates and was selected for the surveillance program. The surveillance weld was fabricated with weld wire Heat

13253 and Linde 1092 flux, Lot # 3833/3774 utilizing the same fabrication practice as that used to fabricate the actual vessel beltline welds. This weld was made of one of the heats and with the same type flux as was used in the intermediate shell longitudinal weld seams.

Thus, the Salem Unit 2 surveillance program meets the intent of this criterion.

Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

Salem Generating Station Unit 2 PTLR Revision 0 Page 36 of 40 The surveillance capsule analysis report, WCAP-15692 [Ref. C-5], was reviewed and it was determined that this criterion is met, consistent with the analysis in these documents.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. C-4].

The functional form of the least-squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these RTNDT values about this line is less than 28°F for welds and less than 17°F for the plates.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998 [Ref. C-3]. At this meeting, the NRC presented five cases. Of the five cases, Case 1 (Surveillance Data Available from Plant but No Other Source) most closely represents the situation for the Salem Unit 2 surveillance plate and weld materials.

Case 5 (Surveillance Data from Other Sources Only) most closely represents the situation for the Salem Unit 2 Lower Shell Longitudinal Weld Seams 3-442 A, 3-442 B, and 3-442 C (Heat

21935/12008) weld material.

Evaluation of the Salem Unit 2 Data Only (Case 1)

Following the NRC Case 1 guidelines, the Salem Unit 2 surveillance plates and weld metal (Heat

13253) will be evaluated using the Salem Unit 2 data. Table C-2 provides the calculation of the

Salem Generating Station Unit 2 PTLR Revision 0 Page 37 of 40 interim CF for Salem Unit 2. Note that when evaluating the credibility of the surveillance weld data, the measured RTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2 Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. In addition, only Salem Unit 2 data are being considered; therefore, no temperature adjustment is required.

Table C-2 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Salem Unit 2 Surveillance Capsule Data Only Capsule Fluence(a) RTNDT(a) FF*RTNDT Material Capsule FF(b) FF2 (x 1019 n/cm2, (°F) (°F)

E > 1.0 MeV)

T 0.273 0.646 61.66 39.84 0.418 Intermediate Shell U 0.581 0.848 66.54 56.43 0.719 Plate B4712-2 (longitudinal) X 1.13 1.034 93.82 97.02 1.069 Y 1.83 1.166 105.69 123.21 1.359 T 0.273 0.646 74.83 48.35 0.418 Intermediate Shell U 0.581 0.848 98.26 83.33 0.719 Plate B4712-2 (transverse) X 1.13 1.034 125.15 129.42 1.069 Y 1.83 1.166 129.33 150.76 1.359 SUM: 728.36 7.130 CFB4712-2 = (FF

RTNDT) ÷ (FF2) = (728.36) ÷ (7.130) = 102.2°F T 0.273 0.646 153.17 98.97 0.418 U 0.581 0.848 185.94 157.68 0.719 Salem Unit 2 surveillance weld X 1.13 1.034 195.43 202.10 1.069 material Y 1.83 1.166 200.90 234.20 1.359 (Heat # 13253)

SUM: 692.95 3.565 CFSurv. Weld = (FF

RTNDT) ÷ (FF2) = (692.95) ÷ (3.565) = 194.4°F Notes:

(a) Fluence and measured RTNDT taken from Table 4-1 of [2].

(b) FF = fluence factor = f(0.28 - 0.10l*log(f)).

Salem Generating Station Unit 2 PTLR Revision 0 Page 38 of 40 The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2 Position 2.1 is presented in Table C-3.

Table C-3 Surveillance Capsule Data Scatter about the Best-Fit Line Using Only Salem Unit 2 Surveillance Data Capsule < 17°F CF(a) Measured Predicted Scatter Fluence(b) (Base Metal)

Material Capsule (Slopebest-fit) FF(c) RTNDT(b) RTNDT(d) RTNDT(e)

(x 1019 n/cm2, < 28°F

(°F) (°F) (°F) (°F)

E > 1.0 MeV) (Weld)

T 102.2 0.273 0.646 61.66 66.0 4.4 Yes Intermediate Shell Plate U 102.2 0.581 0.848 66.54 86.6 20.1 No B4712-2 X 102.2 1.13 1.034 93.82 105.6 11.8 Yes (longitudinal)

Y 102.2 1.83 1.166 105.69 119.1 13.4 Yes T 102.2 0.273 0.646 74.83 66.0 8.8 Yes Intermediate Shell Plate U 102.2 0.581 0.848 98.26 86.6 11.6 Yes B4712-2 X 102.2 1.13 1.034 125.15 105.6 19.5 No (transverse)

Y 102.2 1.83 1.166 129.33 119.1 10.2 Yes T 194.4 0.273 0.646 153.17 125.6 27.6 Yes Salem Unit 2 surveillance U 194.4 0.581 0.848 185.94 164.8 21.1 Yes weld material X 194.4 1.13 1.034 195.43 201.0 5.6 Yes (Heat # 13253)

Y 194.4 1.83 1.166 200.90 226.6 25.7 Yes Notes:

(a) CFs calculated in Table C-2.

(b) Fluence and measured RTNDT values are taken from Table 5-1 of [2].

(c) FF = fluence factor = f(0.28 - 0.10*log (f)).

(d) Predicted RTNDT = CF x FF.

(e) Scatter RTNDT = Absolute Value [Predicted RTNDT - Measured RTNDT].

From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table C-3 indicates that six of the eight (75%) surveillance data points fall inside the +/- 1 of 17F scatter band for surveillance base metals. Therefore, all the plate data are deemed credible per the third criterion.

Table C-3 indicates that all surveillance data points fall inside the +/- 1 of 28F scatter band for surveillance weld materials. 100% of the data are bounded; therefore, the surveillance weld data

Salem Generating Station Unit 2 PTLR Revision 0 Page 39 of 40 are deemed credible per the third criterion. Note that, while credible, the data for surveillance weld Heat # 13253 was determined not to be relevant to the welds in the reactor vessel; therefore, the data are not used for embrittlement calculations of the Salem Unit 2 reactor vessel.

Evaluation of Weld Data from All Sources (Case 5)

Next, data from all sources are considered in order to evaluate the credibility of Heat #

21935/12008 used in the Salem Unit 2 Lower Shell Longitudinal Weld Seams 3-442 A, 3-442 B, and 3-442 C using the NRC Case 5 guidelines.

WCAP-17315-NP [Ref. C-6] already concluded that the data for the Diablo Canyon Unit 2 surveillance weld are credible, therefore, the data are not re-analyzed separately herein.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.

The capsule specimens are located in the reactor between the core barrel and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pad. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F. Hence, this criterion is met.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The Salem Unit 2 surveillance program does not include correlation monitor material. Therefore, this criterion is not applicable to Salem Unit 2.

Salem Generating Station Unit 2 PTLR Revision 0 Page 40 of 40 CONCLUSION:

Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2 [Ref.

C-1], Section B, the Salem Unit 2 surveillance data are credible.

APPENDIX C

REFERENCES:

C-1 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.

Nuclear Regulatory Commission, May 1988. [ADAMS Accession Number ML003740284]

C-2 Code of Federal Regulations 10 CFR 50, Appendix G, Fracture Toughness Requirements, U.S.

Nuclear Regulatory Commission, Federal Register, November 29, 2019.

C-3 K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC Presentation, Generic Letter 92-01 and RPV Integrity Assessment, Status, Schedule, and Issues, NRC/Industry Workshop on RPV Integrity Issues, February 1998. [ADAMS Accession Number ML110070570]

C-4 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society for Testing and Materials, 1982.

C-5 Westinghouse Report WCAP-15692, Revision 0, Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, August 2001.

C-6 Westinghouse Report WCAP-17315-NP, Revision 0, Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations, July 2011.

LR-N22-0012 LAR S22-01 Enclosure 3 Westinghouse WCAP-16982-NP, Rev. 3, Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 WCAP-16982-NP December 2020 Revision 3 Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16982-NP Revision 3 Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity D. Brett Lynch*

RV/CV Design & Analysis December 2020 Reviewer: Benjamin E. Mays*

License Renewal, Radiation Analysis, and Nuclear Operations Approved: Lynn A. Patterson*, Manager RV/CV Design & Analysis

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

Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

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Westinghouse Non-Proprietary Class 3 iii RECORD OF REVISION Revision 0: Original Issue Revision 1: This revision incorporates instrument uncertainties into the Leak Test curves and updates the Pressure-Temperature (P-T) limit curves to account for vacuum refill.

Revision 2: This revision corrects a typographical error on the x-axis of Figures 6-1 and 6-2.

Revision 3: This revision updates the minimum criticality temperature in Figure 6-1 and Table 6-2 to be consistent with the minimum temperature for the leak test in Table 6-2 (289°F).

Changes are marked with revision bars.

WCAP-16982-NP December 2020 Revision 3

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iv Westinghouse Non-Proprietary Class 3 TABLE OF CONTENTS TABLE OF CONTENTS ....................................................................................................................................... iv LIST OF TABLES .................................................................................................................................................. v LIST OF FIGURES ............................................................................................................................................... vi EXECUTIVE

SUMMARY

................................................................................................................................... vii 1 TIME LIMITED AGING ANALYSIS ......................................................................................................1-1 2 CALCULATED BELTLINE FLUENCE ..................................................................................................2-1 3 FRACTURE TOUGHNESS PROPERTIES .............................................................................................3-1 4 PRESSURIZED THERMAL SHOCK ......................................................................................................4-1 5 UPPER SHELF ENERGY ........................................................................................................................5-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ...................................6-1 6.1 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS ............................................... 6-2 6.2 MINIMUM BOLTUP TEMPERATURE ............................................................................ 6-2 6.3 LTOP SYSTEM ENABLE TEMPERATURE .................................................................... 6-3 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE .................................................................7-1 8 ERG P-T LIMITS CATEGORIZATION ...................................................................................................8-1 9 REFERENCES ..........................................................................................................................................9-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIT) ......................................................... A-1 WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 v LIST OF TABLES Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 ............... 1-2 Table 2-1 Salem Unit 2 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface [n/cm2, (E > 1.0 MeV)]................................. 2-2 Table 2-2 Summary of Salem Unit 2 RPV Extended Beltline Materials ......................................... 2-3 Table 3-1 Best Estimate Cu and Ni Weight Percent, Initial RTNDT Values and Initial USE Values for the Salem Unit 2 RPV Beltline and Extended Beltline Materials .................. 3-2 Table 3-2 Calculated Integrated Neutron Exposure of the Surveillance Capsules at Salem Unit 2 ............................................................................................................................... 3-4 Table 3-3 Calculation of CF Values using Salem Unit 2 Surveillance Capsule Test Results........... 3-5 Table 3-4 Summary of the Salem Unit 2 RPV Beltline and Extended Beltline Material Chemistry Factors based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 ...................................................................................................................... 3-6 Table 4-1 Calculation of RTPTS Values for 50 EFPY at the Clad/Base Metal Interface ................... 4-2 Table 5-1 Predicted Position 1.2 USE Values at 50 EFPY (EOLR) ................................................. 5-2 Table 5-2 Predicted Position 2.2 USE Values at 50 EFPY (EOLR) ................................................. 5-3 Table 6-1 Summary of the Limiting ART Values used in Generation of the Salem Unit 2 Reactor Vessel Heatup and Cooldown Curves ................................................................. 6-1 Table 6-2 50 EFPY Heatup Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature] ....... 6-6 Table 6-3 50 EFPY Cooldown Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature] ....... 6-8 Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule............................................ 7-1 Table 8-1 ERG P-T Limits Categories ............................................................................................. 8-1 Table A-1 KIT Values for 100°F/hr. Heatup Curve for 50 EFPY ..................................................... A-2 Table A-2 KIT Values for 100°F/hr. Cooldown Curve for 50 EFPY ................................................ A-3 WCAP-16982-NP December 2020 Revision 3

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vi Westinghouse Non-Proprietary Class 3 LIST OF FIGURES Figure 5-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper Shelf Energy as a Function of Copper and Fluence ................................................................................... 5-4 Figure 6-1 Salem Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60°F/hr. and 100°F/hr.) Applicable for 50 EFPY (with the Flange-Notch &

with Margins for Instrumentation Errors) Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology (w/KIC) .......................................................................................... 6-4 Figure 6-2 Salem Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr.) Applicable for 50 EFPY (with the Flange-Notch & with Margins for Instrumentation Errors) Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology (w/KIC) .......................................................................................... 6-5 WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 vii EXECUTIVE

SUMMARY

This report presents the Time Limited Aging Analyses for the Salem Unit 2 reactor pressure vessel in accordance with the requirements of the License Renewal Rule, 10 CFR Part 54. Time Limited Aging Analyses are calculations which evaluate some safety-related aspects of the reactor pressure vessel within the bounds of the current 40-year license that must be re-done to account for an extended period of operation.

A summary of results for the Salem Unit 2 TLAA is provided below. Based on the TLAA results, it is concluded that the Salem Unit 2 reactor vessel will remain adequate through the extended period of operation.

Identification of Extended Beltline Materials The additional or new vessel materials that exceed 1.0E+17 n/cm2 (E > 1.0 MeV) fluence at End-Of-License Renewal (EOLR) that were not considered in previous evaluations are the following:

Upper Shell Plates B4711-1, B4711-2, and B4711-3

Upper Shell Longitudinal Weld Seams 1-442 A, B, and C

Intermediate Shell to Upper Shell Circumferential Weld Seam 8-442 These are considered to be the extended beltline materials. See Section 2 for more details.

EOLR PTS Values All of the Salem Unit 2 reactor vessel materials that exceed a surface fluence of 1.0E+17 n/cm2 (E > 1.0 MeV) at 50 EFPY are below the RTPTS screening criteria values of 270°F, for axially oriented welds and plates\forgings, and 300°F, for circumferentially oriented welds, at 50 EFPY.

See Section 4 for more details.

EOLR USE Values All of the beltline materials in the Salem Unit 2 reactor vessel are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50 Appendix G) at 50 EFPY. See Section 5 for more details.

EOLR ART Values The Lower Shell Longitudinal Weld Seams 3-442 A&C resulted in the highest ART value at the 1/4 vessel thickness (T) and the 3/4T locations. These limiting ART values are summarized below in Table 6-1 of Section 6. See Section 6 for more details.

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viii Westinghouse Non-Proprietary Class 3 Heatup and Cooldown P-T Limit Curves Applicable to EOLR New heatup and cooldown P-T limit curves, applicable to 50 EFPY, were developed based on the limiting 1/4T and 3/4T ART values shown in Table 6-1 of Section 6. The resulting heatup and cooldown P-T limit curves are shown in Figures 6-1 and 6-2, respectively, with the corresponding data points shown in Tables 6-2 and 6-3. See Section 6 for more details.

Furthermore, Appendix A provides the thermal stress intensity factors (KIT) for the maximum heatup and cooldown rates at 50 EFPY in Tables A-1 and A-2, respectively.

LTOP System Enable Temperature The minimum required LTOP system enable temperature (without margins for temperature uncertainty) for the Salem Unit 2 reactor vessel was conservatively chosen to be 274°F for 50 EFPY. See Section 6 for more details.

Surveillance Capsule Withdrawal Schedule The maximum EOLR shift in transition temperature of the Salem Unit 2 reactor vessel materials that exceed a surface fluence of 1.0E+17 n/cm2 (E > 1.0 MeV) is 229.3°F for 50 EFPY.

Therefore, per Table 1 of ASTM E185-82, the recommended minimum number of surveillance capsules to be withdrawn is five. The new conclusion of a five capsule withdrawal schedule is based on a 60 year license. The new withdrawal schedule is shown in Table 7-1 of Section 7.

See Section 7 for more details.

ERG P-T Limit Categorization The Salem Unit 2 reactor vessel material with the highest RTNDT value is the lower shell plate longitudinal weld seams 3-442 A&C. Thus, this material is considered the limiting material for Salem Unit 2. The lower shell plate longitudinal weld seams 3-442 A&C have an EOLR RTPTS (RTNDT) value of 239°F (based on Table 4-1 of Section 4). The EOLR RTPTS (RTNDT) value of 239°F places Salem Unit 2 in Category II.

The operating time at which Salem Unit 2 would transition from ERG P-T limit Category I to Category II was determined to be approximately 24.6 EFPY. See Section 8 for more details.

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Westinghouse Non-Proprietary Class 3 1-1 1 TIME LIMITED AGING ANALYSIS Time-limited aging analyses (TLAAs) are those licensee calculations that:

Consider the effects of aging

Involve time-limited assumptions defined by the current operating term, for example, 40 years

Involve systems, structures, and components (SSCs) within the scope of license renewal

Involve conclusions or provide the basis for conclusions related to the capability of the system, structure, or component to perform its intended functions

Were determined to be relevant by the licensee in making a safety determination

Are contained or incorporated by reference in the current licensing basis (CLB)

The potential TLAAs for the reactor pressure vessel (RPV) are identified in Table 1-1 along with indication of whether or not they meet the six criteria of 10 CFR 54.3 [Ref. 1] for TLAAs.

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1-2 Westinghouse Non-Proprietary Class 3 Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 Time-Limited Aging Analysis Calculated Pressurized Upper Pressure Beltline Thermal Shelf Limits for Fluence Shock Energy Heatup and Cooldown Considers the Effects of Aging YES YES YES YES Involves Time-Limited YES YES YES YES Assumptions Defined by the Current Operating Term Involves SSC Within the Scope of YES YES YES YES License Renewal Involves Conclusions or Provides YES YES YES YES the Basis for Conclusions Related to the Capability of SSC to Perform Its Intended Function Determined to be Relevant by the YES YES YES YES Licensee in Making a Safety Determination Contained or Incorporated by YES YES YES YES Reference in the CLB WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED BELTLINE FLUENCE At currently licensed service times and operating conditions, the Salem Unit 2 RPV fracture toughness properties provide adequate margins of safety against vessel failure. However, as a vessel accumulates more and more service time, neutron irradiation (fluence) reduces material fracture toughness and initial safety margins. Prevention of RPV failure depends primarily on maintaining RPV material fracture toughness at levels that resist brittle fracture during plant operation. The first step in the TLAA of vessel embrittlement is the calculation of the neutron fluence that causes the embrittlement to increase with time.

The reactor vessel beltline neutron fluence values applicable to a postulated 20-year license renewal period were calculated for each of the Salem Unit 2 RPV beltline materials. The analysis methods used to calculate the Salem Unit 2 vessel fluences satisfy the requirements set forth in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [Ref. 2].

In accordance with Item IV.A2-23 of NUREG-1801, Volume 2, Revision 1 [Ref. 3], any materials exceeding 1.0E+17 n/cm2 (E > 1.0 MeV) must be monitored to evaluate the changes in fracture toughness. RPV materials that are not traditionally thought of as being plant limiting because of low levels of neutron radiation must now be evaluated to determine the accumulated fluence at 50 EFPY. Therefore, fluence calculations were performed for the Salem Unit 2 RPV upper shell and nozzle forgings to determine if they will exceed 1.0E+17 n/cm2 (E > 1.0 MeV) at 50 EFPY. The materials that exceed this threshold are referred to as extended beltline materials in this report and are evaluated to determine their impact to the proposed license renewal period.

In all cases, the maximum exposure occurs at the radial location of the pressure vessel clad/base metal interface. Data is given for the nominal end of Cycle 15 (16.1 Effective Full Power Years (EFPY)) as well as for projections through 50 EFPY of reactor operation. Projections for future operation were based on the continued use of the Cycle 15 spatial power distribution and a core power level of 3459 MWt. Table 2-1 summarizes the maximum projected neutron fluence for each of the reactor pressure vessel beltline region materials. From Table 2-1, it is noted that, although the upper shell course and associated longitudinal welds and the upper shell to intermediate shell circumferential weld are projected to exceed the 1.0E+17 n/cm2 (E > 1.0 MeV) threshold, the nozzles themselves as well as the nozzle to upper shell welds remain below 1.0E+17 n/cm2 (E > 1.0 MeV) through 50 EFPY of operation. Likewise, the lower shell to lower head circumferential weld remains out of the beltline region through 50 EFPY.

The fluence analysis of the upper shell and nozzles revealed that additional materials will exceed the 1.0E+17 n/cm2 (E > 1.0 MeV) threshold. These extended beltline materials, including the respective IDs and heat numbers, are shown in Table 2-2.

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2-2 Westinghouse Non-Proprietary Class 3 Table 2-1 Salem Unit 2 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface [n/cm2, (E > 1.0 MeV)]

Int. Int. Upper Upper Lower Shell Shell Shell Shell Inlet Shell Operating Long. Long. Long. Long. and Nozzle to Lower Time Weld Weld Weld Weld Outlet to Shell Head

[efpy] 0 120/240 60/300 180 Nozzles Welds Weld 16.1 2.86E+18 5.05E+18 8.43E+16 4.78E+16 <1.0E+17 <1.0E+17 <1.0E+17 17.4 3.04E+18 5.42E+18 9.08E+16 5.09E+16 <1.0E+17 <1.0E+17 <1.0E+17 18.8 3.21E+18 5.78E+18 9.74E+16 5.41E+16 <1.0E+17 <1.0E+17 <1.0E+17 20.0 3.37E+18 6.11E+18 1.03E+17 5.69E+16 <1.0E+17 <1.0E+17 <1.0E+17 22.0 3.62E+18 6.64E+18 1.13E+17 6.15E+16 <1.0E+17 <1.0E+17 <1.0E+17 24.0 3.88E+18 7.18E+18 1.23E+17 6.61E+16 <1.0E+17 <1.0E+17 <1.0E+17 26.0 4.14E+18 7.71E+18 1.32E+17 7.07E+16 <1.0E+17 <1.0E+17 <1.0E+17 28.0 4.39E+18 8.25E+18 1.42E+17 7.53E+16 <1.0E+17 <1.0E+17 <1.0E+17 30.0 4.65E+18 8.78E+18 1.51E+17 8.00E+16 <1.0E+17 <1.0E+17 <1.0E+17 32.0 4.91E+18 9.32E+18 1.61E+17 8.46E+16 <1.0E+17 <1.0E+17 <1.0E+17 34.0 5.16E+18 9.85E+18 1.71E+17 8.92E+16 <1.0E+17 <1.0E+17 <1.0E+17 36.0 5.42E+18 1.04E+19 1.80E+17 9.38E+16 <1.0E+17 <1.0E+17 <1.0E+17 38.0 5.68E+18 1.09E+19 1.90E+17 9.85E+16 <1.0E+17 <1.0E+17 <1.0E+17 40.0 5.93E+18 1.15E+19 2.00E+17 1.03E+17 <1.0E+17 <1.0E+17 <1.0E+17 42.0 6.19E+18 1.20E+19 2.09E+17 1.08E+17 <1.0E+17 <1.0E+17 <1.0E+17 44.0 6.45E+18 1.25E+19 2.19E+17 1.12E+17 <1.0E+17 <1.0E+17 <1.0E+17 46.0 6.70E+18 1.31E+19 2.29E+17 1.17E+17 <1.0E+17 <1.0E+17 <1.0E+17 48.0 6.96E+18 1.36E+19 2.38E+17 1.22E+17 <1.0E+17 <1.0E+17 <1.0E+17 50.0 7.22E+18 1.41E+19 2.48E+17 1.26E+17 <1.0E+17 <1.0E+17 <1.0E+17 WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 2-3 Table 2-2 Summary of Salem Unit 2 RPV Extended Beltline Materials Salem Unit 2 Material Material ID Heat Number Upper Shell B4711-1 C4194-1 C4149-2 Upper Shell B4711-2 C4171-1 Upper Shell B4711-3 20291/1P2809 Linde 1092, Lot Upper Shell Longitudinal Weld Seams 1-442 A, B & C 3854 90099, Linde 0091, Intermediate to Upper Shell Circumferential Weld Seam 8-442 Lot 3977 WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan [Ref. 4]. The beltline and extended beltline material properties of the Salem Unit 2 reactor vessel are presented in Table 3-1. Note that the closure head and vessel flanges were included in Table 3-1 as these materials are considered for the minimum temperature requirement of 10 CFR 50, Appendix G

[Ref. 5] for pressure-temperature limits. The minimum temperature pertains to the controlling material, which is either the material in the closure flange or the material in the beltline region with the highest reference temperature.

The chemistry factors (CFs) were calculated using Positions 1.1 and 2.1 of Regulatory Guide 1.99, Rev. 2 [Ref. 6]. Position 1.1 uses Tables 1 and 2 from the Reg. Guide along with the best estimate copper and nickel weight percents, which are presented in Table 3-1 of this report.

Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The calculated fluence values at the surveillance capsule locations are provided in Table 3-2 and are used to determine the CFs in Table 3-3. The capsule fluence values were determined using ENDF/B-VI cross-sections and followed the guidance in Regulatory Guide 1.190 [Ref. 2].

Table 3-4 summarizes the Positions 1.1 and 2.1 CFs determined for the Salem Unit 2 RPV beltline and extended beltline materials.

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3-2 Westinghouse Non-Proprietary Class 3 Table 3-1 Best Estimate Cu and Ni Weight Percent, Initial RTNDT Values and Initial USE Values for the Salem Unit 2 RPV Beltline and Extended Beltline Materials Initial Initial Cu Ni RPV Material Description Material Heat Number USE RTNDT

(%) (%)

(ft-lb) (°F)

Closure Head Flange B4702-1(a) --- --- --- --- -40 Vessel Flange B5001(a) --- --- --- --- 12 Upper Shell B4711-1(b) C4194-1 0.11 0.55 87.1(c) 60 (b) (d)

Upper Shell B4711-2 C4149-2 0.14 0.56 79.3 60 Upper Shell B4711-3(b) C4171-1 0.12 0.58 69.3(e) 101 Intermediate Shell B4712-1(f) C4173-1 0.13 0.56 106 0 (f)

Intermediate Shell B4712-2 C4186-2 0.12 0.61 97 12 (f)

Intermediate Shell B4712-3 C4194-2 0.11 0.57 107 10 Lower Shell B4713-1(f) C4182-1 0.12 0.60 98 8 (f)

Lower Shell B4713-2 C4182-2 0.12 0.57 103 8 (f)

Lower Shell B4713-3 B-8343-1 0.12 0.58 121 10 Intermediate to Upper Shell 20291/1P2809 0.27(g) 0.735(g) 97(h) -56(i)

Circumferential Weld Seam 8-442 Linde 1092, Lot 3854 Intermediate to Lower Shell 90099, 0.197 0.060 99.7 -56(i)

Circumferential Weld Seam 9-442(f) Linde 0091, Lot 3977 Upper Shell Longitudinal Weld 13253/12008, 0.21(g) 0.873(g) 97(h) -56(i)

Seams 1-442 A, B & C Linde 1092, Lot 3833 Intermediate Shell Longitudinal Weld 13253/20291, 0.221 0.732 96.2 -56(i)

Seams 2-442 A, B & C(f) Linde 1092, Lot 3833 Lower Shell Longitudinal Weld 21935/12008, 0.213 0.867 114 -56(i)

Seams 3-442 A, B & C(f) Linde 1092, Lot 3889 13253, Surveillance Weld Material(f) 0.225 0.727 --- ---

Linde 1092 Lot 3833/3774 Notes:

(a) The initial RTNDT value for the vessel flange was obtained from WCAP-15693 [Ref. 7]. Due to the reactor vessel head replacement, the closure head flange was also replaced and the new initial RTNDT value was provided in S-TODI-2008-0010 [Ref. 8].

(b) Values from CMTR-RV-PNJ [Ref. 9]. Note that the initial RTNDT values were determined in accordance with the requirements of Subparagraph NB-2331 of Section III of the ASME B&PV Code [Ref. 10], as specified by Paragraph II - D of 10 CFR Part 50, Appendix G [Ref. 5]. These fracture toughness requirements are also summarized in Branch Technical Position MTEB Section II.5-2 (Fracture Toughness) of the NRC Regulatory Standard Review Plan [Ref. 4]. Since transversely oriented Charpy V-Notch specimens were not tested for the upper shell plates, the initial RTNDT was determined using Section 1.1(3)(a) and (b).

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Westinghouse Non-Proprietary Class 3 3-3 (c) Average USE value calculated based on ASTM E185-82 [Ref. 11] and Charpy test data points documented in CMTR-RV-PNJ [Ref. 9]: 133 and 135 ft-lbs. Due to specimen orientation, this average value has been conservatively reduced by 65%.

(d) Average USE value calculated based on ASTM E185-82 [Ref. 11] and Charpy test data points documented in CMTR-RV-PNJ [Ref. 9]: 123,125, and 118 ft-lbs. Due to specimen orientation, this average value has been conservatively reduced by 65%.

(e) Average USE value calculated based on ASTM E185-82 [Ref. 11] and Charpy test data points documented in CMTR-RV-PNJ [Ref. 9]: 106,105, and 109 ft-lbs. Due to specimen orientation, this average value has been conservatively reduced by 65%.

(f) Values from WCAP-15693 [Ref. 7].

(g) Best estimate copper and nickel content based on CE NPSD-1119, Revision 1 [Ref. 12]. Note that in the case of Intermediate to Upper Shell Circumferential Weld Seam 8-442, there is no record that the two heats of 20291 and 1P2809 were deposited in tandem; therefore, the copper and nickel content for heat 1P2809 is used as a more conservative basis for estimating the effect of neutron irradiation.

(h) Generic upper shelf energy values are based on CEN-622-A [Ref. 13].

(i) Generic initial RTNDT values per 10 CFR 50.61 [Ref. 14].

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3-4 Westinghouse Non-Proprietary Class 3 Table 3-2 Calculated Integrated Neutron Exposure of the Surveillance Capsules at Salem Unit 2 Capsule Fluence T 2.75E+18 n/cm2, (E > 1.0 MeV)

U 5.82E+18 n/cm2, (E > 1.0 MeV)

X 1.12E+19 n/cm2, (E > 1.0 MeV)

Y 1.82E+19 n/cm2, (E > 1.0 MeV)

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Westinghouse Non-Proprietary Class 3 3-5 Table 3-3 Calculation of CF Values using Salem Unit 2 Surveillance Capsule Test Results Capsule RPV Material Capsule FF(b) RTNDT(c) FF*RTNDT FF2 Fluence(a)

(E+19 n/cm2) °F °F Intermediate Shell T 0.275 0.648 61.66 39.96 0.420 B4712-2 U 0.582 0.849 66.54 56.46 0.720 (Longitudinal) (c) X 1.12 1.032 93.82 96.79 1.064 Y 1.82 1.164 105.69 123.05 1.356 T 0.275 0.648 74.83 48.49 0.420 Intermediate Shell U 0.582 0.849 98.26 83.37 0.720 B4712-2 X 1.12 1.032 125.15 129.11 1.064 (Transverse) (c)

Y 1.82 1.164 129.33 150.58 1.356 Sum = 727.81 7.119 CF = (FF

RTNDT) / (FF2) = (727.81 / 7.119) = 102.2°F T 0.275 0.648 153.17 99.25 0.420 U 0.582 0.849 185.94 157.77 0.720 X 1.12 1.032 195.43 201.62 1.064 Surveillance Weld Material(d) Y 1.82 1.164 200.90 233.90 1.356 Sum = 692.55 3.560 CF = (FF

RTNDT) / (FF2) = (692.55 / 3.560) = 194.5°F Notes:

(a) f = calculated fluence values from Table 3-2.

(b) FF = fluence factor = f(0.28-0.1*log(f)).

(c) RTNDT values are the measured 30 ft-lb shift values from Table 4-6 of WCAP-15693 [Ref. 7].

(d) Since the surveillance weld metal is only representative of the beltline welds, and not identical, it is not used in any of the calculations documented in this report. Thus, there is no Position 2.1 CF value available for the beltline welds.

This is consistent with WCAP-15693 [Ref. 7].

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3-6 Westinghouse Non-Proprietary Class 3 Table 3-4 Summary of the Salem Unit 2 RPV Beltline and Extended Beltline Material Chemistry Factors based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Chemistry Factor (°F)

RPV Material Position 1.1 Position 2.1 Upper Shell B4711-1 73.5 ---

Upper Shell B4711-2 98.2 ---

Upper Shell B4711-3 82.6 ---

Intermediate Shell B4712-1 89.8 ---

Intermediate Shell B4712-2 83.2 102.2 Intermediate Shell B4712-3 73.7 ---

Lower Shell B4713-1 83.0 ---

Lower Shell B4713-2 82.4 ---

Lower Shell B4713-3 82.6 ---

Intermediate to Upper Shell Circumferential Weld Seam 8-442 205.6 ---

(Heat # 1P2809)

Intermediate to Lower Shell Circumferential Weld Seam 9-442 91.4 ---

(Heat # 90099)

Upper Shell Longitudinal Weld Seams 1-442 A, B & C 208.7 ---

(Heat # 13253/12008)

Intermediate Shell Longitudinal Weld Seams 2-442 A, B & C 189.1 ---

(Heat # 13253/20291)

Lower Shell Longitudinal Weld Seams 3-442 A, B & C 208.6 ---

(Heat # 21935/12008)

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Westinghouse Non-Proprietary Class 3 4-1 4 PRESSURIZED THERMAL SHOCK A limiting condition on RPV integrity known as Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break.

Such transients may challenge the integrity of the RPV under the following conditions: severe overcooling of the inside surface of the vessel wall followed by high repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect anywhere within the vessel wall.

In 1985, the U.S. NRC issued a formal ruling (10 CFR 50.61) on PTS [Ref. 14] that established screening criteria on PWR vessel embrittlement, as measured by the maximum nil ductility reference temperature in the limiting beltline component, termed RTPTS. RTPTS screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end of license. The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RTPTS) values consistent with the methods given in Regulatory Guide 1.99, Revision 2 [Ref. 6].

These accepted methods were used with the surface fluence of Section 2 to calculate the following RTPTS values for the Salem Unit 2 RPV materials at 50 EFPY, which is the end of the license renewal period (see Table 4-1).

PTS Conclusion The limiting RTPTS value for the axially oriented welds and plates is 239°F (see Table 4-1); this value corresponds to the Lower Shell Longitudinal Weld Seams 3-442 A&C.

The limiting RTPTS value for the circumferentially oriented welds is 118°F (see Table 4-1); this value corresponds to the Intermediate Shell to Lower Shell Circumferential Weld Seam 9-442.

Therefore, all of the Salem Unit 2 reactor vessel materials that exceed a surface fluence of 1.0E+17 n/cm2 (E > 1.0 MeV) at 50 EFPY are below the RTPTS screening criteria values of 270°F, for axially oriented welds and plates/forgings, and 300°F, for circumferentially oriented welds, at 50 EFPY.

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4-2 Westinghouse Non-Proprietary Class 3 Table 4-1 Calculation of RTPTS Values for 50 EFPY at the Clad/Base Metal Interface Fluence(b)

RPV Material CF(a) RTPTS(d) RTNDT(U)(e) U (g) M RTPTS (E+19 FF(c)

(°F) (°F) (°F) (°F) (°F) (°F) (°F) n/cm2)

Upper Shell B4711-1 73.5 0.0341 0.237 17.39 60 0 8.69 17.39 95 Upper Shell B4711-2 98.2 0.0341 0.237 23.23 60 0 11.61 23.23 106 Upper Shell B4711-3 82.6 0.0341 0.237 19.54 101 0 9.77 19.54 140 Intermediate Shell B4712-1 89.8 1.95 1.182 106.19 0 0 17 34 140 Intermediate Shell B4712-2 83.2 1.95 1.182 98.38 12 0 17 34 144 (without credible surveillance data)

Intermediate Shell B4712-2 102.2 1.95 1.182 120.85 12 0 8.5 17 150 (with credible surveillance data)

Intermediate Shell B4712-3 73.7 1.95 1.182 87.15 10 0 17 34 131 Lower Shell B4713-1 83.0 1.96 1.184 98.26 8 0 17 34 140 Lower Shell B4713-2 82.4 1.96 1.184 97.55 8 0 17 34 140 Lower Shell B4713-3 82.6 1.96 1.184 97.78 10 0 17 34 142 Intermediate to Upper Shell Circumferential Weld Seam 8-442 205.6 0.0341 0.237 48.63 -56(f) 17 24.32 59.34 52 (Heat # 20291/1P2809)

Intermediate to Lower Shell Circumferential Weld Seam 9-442 91.4 1.96 1.184 108.20 -56(f) 17 28 65.51 118 (Heat # 90099)

Upper Shell Longitudinal Weld Seams 1-442 A&C 208.7 0.0248 0.196 40.94 -56(f) 17 20.47 53.22 38 (Heat # 13253/12008)

Upper Shell Longitudinal Weld Seam 1-442 B 208.7 0.0126 0.128 26.72 -56(f) 17 13.36 43.24 14 (Heat # 13253/12008)

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Westinghouse Non-Proprietary Class 3 4-3 Table 4-1 Calculation of RTPTS Values for 50 EFPY at the Clad/Base Metal Interface Fluence(b)

RPV Material CF(a) RTPTS(d) RTNDT(U)(e) U (g) M RTPTS (E+19 FF(c)

(°F) (°F) (°F) (°F) (°F) (°F) (°F) n/cm2)

Intermediate Shell Longitudinal Weld Seam 2-442 A 189.1 0.722 0.909 171.82 -56(f) 17 28 65.51 181 (Heat # 13253/20291)

Intermediate Shell Longitudinal Weld Seam 2-442 B&C 189.1 1.41 1.095 207.13 -56(f) 17 28 65.51 217 (Heat # 13253/20291)

Lower Shell Longitudinal Weld Seams 3-442 A&C 208.6 1.43 1.099 229.30 -56(f) 17 28 65.51 239 (Heat # 21935/12008)

Lower Shell Longitudinal Weld Seam 3-442 B 208.6 0.727 0.911 189.94 -56(f) 17 28 65.51 199 (Heat # 21935/12008)

Notes:

(a) From Table 3-4 of this report.

(b) From Table 2-1 of this report.

(c) FF = fluence factor = f(0.28-0.1*log(f)).

(d) RTPTS = CF

FF.

(e) From Table 3-1 of this report. All RTNDT(U) values are measured values unless otherwise noted. Note that U = 0°F for measured values.

(f) RTNDT(U) values are generic mean values per 10 CFR 50.61. Note that U = 17°F for generic values.

(g) Per WCAP-15692, Rev. 0 [Ref. 15], the surveillance plate and weld data obtained for Salem Unit 2 has been deemed credible. Per the guidance of Reg.

Guide 1.99, Revision 2 [Ref. 6], the base metal = 17°F for Position 1.1 and = 8.5°F for Position 2.1; and the weld metal = 28°F for Position 1.1 and = 14°F for Position 2.1. However, need not exceed 0.5*RTNDT. Note that the Position 2.1 is not used for the surveillance weld data as the surveillance weld is not identical to the beltline welds (this is consistent with WCAP-15693 [Ref. 7]).

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Westinghouse Non-Proprietary Class 3 5-1 5 UPPER SHELF ENERGY The decrease in upper shelf Charpy energy is associated with the determination of acceptable RPV toughness during the license renewal period when the vessel is exposed to additional irradiation.

The requirements on upper shelf energy are included in 10 CFR 50, Appendix G [Ref. 5]. 10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the upper shelf energy of any of the RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.

There are two methods that can be used to estimate the change in upper shelf energy (USE) with irradiation, depending on the availability of credible surveillance capsule data as defined in Regulatory Guide 1.99, Revision 2. For vessel beltline materials that are not in the surveillance program or not credible, the Charpy USE is assumed to decrease as a function of fluence and copper content, as indicated in Regulatory Guide 1.99, Revision 2 [Ref. 6].

When two or more credible surveillance sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material. The surveillance data are then used in conjunction with the Regulatory Guide data to predict the change in USE of the RPV due to irradiation.

Using the 1/4T fluence values, projected upper shelf energy values were calculated to determine if the Salem Unit 2 beltline and extended beltline materials remain above the 50 ft-lb limit at 50 EFPY (see Tables 5-1 and 5-2).

USE Conclusion All of the beltline and extended beltline materials in the Salem Unit 2 reactor vessel are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50 Appendix G) at 50 EFPY.

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5-2 Westinghouse Non-Proprietary Class 3 Table 5-1 Predicted Position 1.2 USE Values at 50 EFPY (EOLR)

USE Cu(a) 1/4T Fluence(b) Initial USE(a) USE RPV Material Decrease

(%) (E+19 n/cm2) (ft-lb) (ft-lb)

(%)

Upper Shell B4711-1 0.11 0.020 87.1 9 79 Upper Shell B4711-2 0.14 0.020 79.3 10 71 Upper Shell B4711-3 0.12 0.020 69.3 9 63 Intermediate Shell B4712-1 0.13 1.162 106 25 80 Intermediate Shell B4712-2 0.12 1.162 97 25 73 Intermediate Shell B4712-3 0.11 1.162 107 25 80 Lower Shell B4713-1 0.12 1.168 98 25 74 Lower Shell B4713-2 0.12 1.168 103 25 77 Lower Shell B4713-3 0.12 1.168 121 25 91 Intermediate to Upper Shell Circumferential Weld 0.27 0.020 97 18 80 Seam 8-442 (Heat # 1P2809)

Intermediate to Lower Shell Circumferential Weld Seam 9-442 0.197 1.168 99.7 35 65 (Heat # 90099)

Upper Shell Longitudinal Weld Seams 1-442 A&C 0.21 0.015 97 16(c) 81 (Heat # 13253/12008)

Upper Shell Longitudinal Weld Seam 1-442 B 0.21 0.008 97 16(c) 81 (Heat # 13253/12008)

Intermediate Shell Longitudinal Weld Seam 2-442 A 0.221 0.430 96.2 33 64 (Heat # 13253/20291)

Intermediate Shell Longitudinal Weld Seam 2-442 B&C 0.221 0.840 96.2 38 60 (Heat # 13253/20291)

Lower Shell Longitudinal Weld Seams 3-442 A&C 0.213 0.852 114 38 71 (Heat # 21935/12008)

Lower Shell Longitudinal Weld Seam 3-442 B 0.213 0.433 114 33 76 (Heat # 21935/12008)

Notes:

(a) From Table 3-1 of this report.

(b) 1/4 T EOLR Fluence = fsurf

exp(-0.24*X), where x is the depth (in inches) into the vessel wall from the inner surface; for the 1/4T location, x = 0.25

8.625 inches = 2.156 inches.

(c) The fluence ranges from 2.0E+17 n/cm2 to 6.0E+19 n/cm2 in Regulatory Guide 1.99, Rev. 2, Figure 2; thus, the upper shelf energy decrease was conservatively estimated based on 2.0E+17n/cm2.

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Westinghouse Non-Proprietary Class 3 5-3 The predicted Position 2.2 EOLR USE value is calculated in Table 5-2 for the surveillance material (i.e., Intermediate Shell Plate B4712-2). The reduced plant surveillance data from Table 5-10 of WCAP-15692 [Ref. 15] was plotted on Reg. Guide 1.99, Revision 2, Figure 2 (see Figure 5-1). This data was fitted by drawing a line parallel to the existing lines as the upper bound of all the surveillance data. This reduced line was used instead of the existing lines to determine the Position 2.2 EOLR USE values.

Table 5-2 Predicted Position 2.2 USE Values at 50 EFPY (EOLR)

Unirradiated USE Cu 1/4T Fluence USE RPV Material USE Decrease

(%) (E+19 n/cm2) (ft-lb)

(ft-lb) (%)

Intermediate Shell B4712-2 0.12 1.162 97 16 81 WCAP-16982-NP December 2020 Revision 3

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5-4 Westinghouse Non-Proprietary Class 3 100.0 Upper Limit

% Copper Base Metal Weld Percentage Drop in USE 0.35 0.30 B4712-2 0.30 0.25 0.25 0.20 0.20 0.15 0.15 0.10 10.0 0.10 0.05 1.0 1.00E+17 1.00E+18 1.00E+19 1.00E+20 2

Neutron Fluence, n/cm (E > 1 MeV)

B4712-2 LT B4712-2 TL 1/4T Fluence Figure 5-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper Shelf Energy as a Function of Copper and Fluence WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 6-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES General Design Criterion 14 of Appendix A of 10 CFR Part 50 [Ref. 16], "Reactor Coolant Pressure Boundary," requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested to have an extremely low probability of abnormal leakage or rapid failure and of gross rupture. Likewise, General Design Criterion 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," requires that the reactor coolant pressure boundary be designed with sufficient margin to ensure that when stressed under operating, maintenance, and testing, the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized. To assess the structural integrity of the RPV, General Design Criterion 32, "Inspection of Reactor Coolant Pressure Boundary," requires an appropriate materials surveillance program for the RPV beltline region.

Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil ductility transition temperature) corresponding to the limiting material in the beltline region of the RPV. The most limiting RTNDT of the material in the core (beltline) region of the RPV is determined by using the unirradiated RPV material fracture toughness properties and estimating the irradiation-induced shift (RTNDT).

RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactor's life, RTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. Using the adjusted reference temperature (ART) values, pressure-temperature limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G [Ref. 5], as augmented by Appendix G to Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code [Ref. 17].

The 1/4 and 3/4 thickness (1/4T and 3/4T) fluences and material properties were used to determine the limiting material and calculate its pressure-temperature limits at 50 EFPY, the end of the license renewal period being evaluated. The limiting materials were determined from the values of ART at the 1/4T and 3/4T locations and summarized in Table 6-1; Lower Shell Longitudinal Weld Seams 3-442 A and C resulted in the highest ART value at the 1/4T and 3/4T locations. These limiting ART values were calculated for an end of license renewal time of 50 EFPY.

Table 6-1 Summary of the Limiting ART Values used in Generation of the Salem Unit 2 Reactor Vessel Heatup and Cooldown Curves 1/4 T Limiting ART 3/4 T Limiting ART EFPY Lower Shell Longitudinal Weld Seams 3-442 A&C 50 209°F 150°F Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been developed utilizing the 1998 through the 2000 Summer Addenda Edition of the ASME Code Section XI, Appendix G methodology along with ASME Code Case N-641 WCAP-16982-NP December 2020 Revision 3

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6-2 Westinghouse Non-Proprietary Class 3

[Ref. 17]. The 1998 through the 2000 Summer Addenda Edition of the ASME Code Section XI, Appendix G methodology allows use of the less restrictive KIC stress intensity factors and also allows the use of the less restrictive Circ-Flaw methodology (formerly known as ASME Code Cases N-640 and N-588, respectively). Code Case N-641 provides alternative procedures for calculating the allowable pressure-temperature relationships and Low Temperature Overpressure Protection (LTOP) effective temperatures. Code Case N-641 broke Section 2215 of the 1998 through the 2000 Summer Addenda Edition of Section XI, Appendix G down into Sections 2215.1 and 2215.2 for the allowable pressures and LTOP System, respectively. Section 2215.1 of Code Case N-641 replaced all KIA designations with KIC, thus removing the option to use the more restrictive KIA stress intensity factor. Section 2215.2 provided the methodology to determine the LTOP System effective temperature. These methodologies are documented in WCAP-14040-A, Revision 4 [Ref. 18].

Figure 6-1 presents the limiting heatup curves with margins for possible instrumentation errors

[61 psi and 18°F] using heatup rates of 60 and 100°F/hr. applicable for 50 EFPY with the Flange-Notch requirement. Figure 6-2 presents the limiting cooldown curves with margins for possible instrumentation errors [61 psi and 18°F] using cooldown rates of 0, 20, 40, 60 and 100°F/hr. applicable for 50 EFPY with the Flange-Notch requirement. The corresponding data points for Figures 6-1 and 6-2 are provided in Tables 6-2 and 6-3.

6.1 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Ref. 5] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation and 90°F for hydrostatic pressure tests and leak tests when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure (3106 psig), which is 621 psig for Salem Unit 2. The limiting unirradiated RTNDT of 12°F (due to limitations in the OPERLIM code, this temperature was conservatively rounded to the next multiple of five, 15°F) occurs in the vessel flange of the Salem Unit 2 reactor vessel, so the minimum allowable temperature of this region is 153°F at pressures greater than 560 psig (with instrument uncertainties). These limits are shown in Figure 6-1 and 6-2.

6.2 MINIMUM BOLTUP TEMPERATURE According to WCAP-14040-A, Rev. 4 [Ref. 18], the minimum boltup temperature should be 60°F or the material RTNDT of the stressed region, whichever is higher. Based on the initial RTNDT values for the closure head and vessel flange in Table 3-1, the minimum boltup temperature for Salem Unit 2 is 60°F. Per S-TODI-2008-0010 [Ref. 8], a margin of 2°F was applied to the minimum boltup temperature. Thus, the minimum boltup temperature with margin is 62°F.

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Westinghouse Non-Proprietary Class 3 6-3 6.3 LTOP SYSTEM ENABLE TEMPERATURE The minimum LTOP System enable temperature was determined to be 274°F, utilizing the methodology of ASME Code Case N-641.

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6-4 Westinghouse Non-Proprietary Class 3 MATERIAL PROPERTY BASIS LIMITING MATERIAL AND ART VALUES AT 50 EFPY:

1/4T ART: 209°F (Lower Shell Longitudinal Weld Seams 3-442 A&C) 3/4T ART: 150°F (Lower Shell Longitudinal Weld Seams 3-442 A&C)

Operlim Version:5.2 Run:4802 Operlim.xls Version: 5.2 2500 Leak Test Limit 2250 2000 Unacceptable Acceptable Operation Operation 1750 1500 Calculated Pressure (PSIG) 1250 Critical Limit 60 Deg. F/Hr 1000 Critical Limit 100 Deg. F/Hr 750 Heatup Rate 60 Deg. F/Hr Criticality Limit based on 500 Heatup Rate inservice hydrostatic test 100 Deg. F/Hr temperature (289°F) for the service period up to 50 EFPY 250 Boltup Temperature = 62°F (includes 2°F margin)

Lower limit for RCS pressure is 0 psia without uncertainty 0

0 50 100 150 200 200 250 250 300 300 350 350 400 400 450 450 500 500 550 550 Moderator Temperature (Deg. F)

-250 Figure 6-1 Salem Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60°F/hr.

and 100°F/hr.) Applicable for 50 EFPY (with the Flange-Notch & with Margins for Instrumentation Errors) Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology (w/KIC)

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Westinghouse Non-Proprietary Class 3 6-5 MATERIAL PROPERTY BASIS LIMITING MATERIAL AND ART VALUES AT 50 EFPY:

1/4T ART: 209°F (Lower Shell Longitudinal Weld Seams 3-442 A&C) 3/4T ART: 150°F (Lower Shell Longitudinal Weld Seams 3-442 A&C) 2500 Operlim Version:5.2 Run:4802 Operlim.xls Version: 5.2 2250 2000 Unacceptable Acceptable 1750 Operation Operation 1500 Calculated Pressure (PSIG) 1250 1000 Cooldown Rates 0°F/hr. (steady-state)

-20°F/hr.

-40°F/hr.

750 -60°F/hr.

-100°F/hr.

500 250 Boltup Temperature = 62°F (includes 2°F margin)

Lower limit for RCS pressure is 0 psia without uncertainty 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

-250 Figure 6-2 Salem Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr.) Applicable for 50 EFPY (with the Flange-Notch & with Margins for Instrumentation Errors) Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology (w/KIC)

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6-6 Westinghouse Non-Proprietary Class 3 Table 6-2 50 EFPY Heatup Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature]

Leak Test 60°F/hr 60°F/hr 100°F/hr 100°F/hr Limit Heatup Criticality Heatup Criticality T P T P T P (psig) T P T P

(°F) (psig) (°F) (psig) (°F) (°F) (psig) (°F) (psig) 272 2000 62 Note (a) 289 Note (a) 62 Note (a) 289 Note (a) 272 2000 62 555 289 555 62 510 289 510 289 2485 83 555 289 555 83 510 289 511 289 2485 88 555 289 556 88 510 289 511 93 555 289 558 93 510 289 513 98 555 289 560 98 510 289 514 103 555 289 560 103 510 289 516 108 555 289 560 108 510 289 518 113 555 289 560 113 510 289 521 118 556 289 560 118 510 289 523 123 560 289 560 123 510 289 528 128 560 289 560 128 510 289 531 133 560 289 560 133 510 289 536 138 560 289 560 138 511 289 541 143 560 289 560 143 513 289 546 148 560 289 560 148 516 289 552 153 560 289 613 153 521 289 557 153 560 289 627 153 521 289 560 153 613 289 644 153 521 289 560 158 627 289 663 158 528 289 560 163 644 289 684 163 536 289 560 168 662 289 707 168 546 289 570 173 675 289 733 173 557 289 585 178 688 289 762 178 570 289 603 183 703 289 794 183 585 289 622 188 720 289 829 188 603 289 644 193 738 289 868 193 622 289 669 198 758 289 906 198 644 289 696 203 781 289 927 203 669 289 727 208 805 289 950 208 696 289 761 213 833 289 976 213 727 289 798 218 863 289 1003 218 761 289 840 223 896 289 1034 223 798 289 886 228 933 289 1068 228 840 289 938 233 973 289 1105 233 886 289 994 238 1018 289 1132 238 938 289 1067 243 1068 293 1184 243 994 293 1126 248 1123 298 1241 248 1057 298 1203 WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 6-7 Table 6-2 (continued) 50 EFPY Heatup Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi)

Uncertainties, and With 2°F Margin on the Boltup Temperature]

Leak Test 60°F/hr 60°F/hr 100°F/hr 100°F/hr Limit Heatup Criticality Heatup Criticality T P T P T P T P T P

(°F) (psig) (°F) (psig) (°F) (psig) (°F) (psig) (°F) (psig) 253 1184 303 1296 253 1126 303 1287 258 1241 308 1356 258 1203 308 1354 263 1296 313 1423 263 1287 313 1409 268 1356 318 1496 268 1354 318 1470 273 1423 323 1577 273 1409 323 1537 278 1496 328 1667 278 1470 328 1611 283 1577 333 1766 283 1537 333 1692 288 1667 338 1875 288 1611 338 1782 293 1766 343 1995 293 1692 343 1880 298 1875 348 2128 298 1782 348 1989 303 1995 353 2274 303 1880 353 2109 308 2128 308 1989 358 2241 313 2274 313 2109 363 2386 318 2241 323 2386 Note:

(a) The lower limit for reactor coolant system pressure is 0 psia without uncertainty.

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6-8 Westinghouse Non-Proprietary Class 3 Table 6-3 50 EFPY Cooldown Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi) Uncertainties, and With 2°F Margin on the Boltup Temperature]

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

T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 62 Note (a) 62 Note (a) 62 Note (a) 62 Note (a) 62 Note (a) 62 560 62 515 62 464 62 411 62 302 83 560 83 517 83 466 83 413 83 304 88 560 88 519 88 468 88 415 88 306 93 560 93 522 93 470 93 418 93 309 98 560 98 525 98 473 98 421 98 313 103 560 103 528 103 476 103 424 103 316 108 560 108 531 108 480 108 428 108 321 113 560 113 535 113 484 113 432 113 326 118 560 118 539 118 489 118 437 118 331 123 560 123 544 123 494 123 442 123 338 128 560 128 549 128 499 128 449 128 345 133 560 133 555 133 506 133 455 133 353 138 560 138 560 138 513 138 463 138 362 143 560 143 560 143 520 143 471 143 372 148 560 148 560 148 529 148 481 148 383 153 560 153 560 153 539 153 491 153 396 153 560 153 560 158 549 158 503 158 410 153 632 153 586 163 561 163 516 163 426 158 641 158 595 168 574 168 530 168 444 163 651 163 606 173 589 173 546 173 463 168 662 168 618 178 605 178 564 178 485 173 675 173 632 183 623 183 584 183 510 178 688 178 646 188 643 188 606 188 537 183 703 183 663 193 665 193 630 193 567 188 720 188 681 198 690 198 658 198 601 193 738 193 701 203 717 203 688 203 638 198 758 198 723 208 747 208 721 208 680 203 781 203 748 213 780 213 758 213 726 208 805 208 775 218 817 218 799 218 777 213 833 213 805 223 858 223 844 223 834 218 863 218 838 228 903 228 895 228 895 223 896 223 875 233 953 233 950 233 950 228 933 228 916 238 1008 238 1008 238 1008 233 973 233 961 243 1066 243 1066 243 1066 238 1018 238 1011 248 1123 248 1123 248 1123 243 1068 243 1066 253 1184 253 1184 253 1184 248 1123 248 1123 258 1251 258 1251 258 1251 253 1184 253 1184 263 1325 263 1325 263 1325 258 1251 258 1251 268 1407 268 1407 268 1407 WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 6-9 Table 6-3 (continued) 50 EFPY Cooldown Curve Data Points Using the 1998 through 2000 Summer Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, App. G Methodology [With KIC, With Flange, With Temperature (18°F) and Pressure (61 psi)

Uncertainties, and With 2°F Margin on the Boltup Temperature]

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

T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 263 1325 263 1325 273 1497 273 1497 273 1497 268 1407 268 1407 278 1597 278 1597 278 1597 273 1497 273 1497 283 1708 283 1708 283 1708 278 1597 278 1597 288 1830 288 1830 288 1830 283 1708 283 1708 293 1965 293 1965 293 1965 288 1830 288 1830 298 2115 298 2115 298 2115 293 1965 293 1965 303 2280 303 2280 303 2280 298 2115 298 2115 303 2280 303 2280 Note:

(a) The lower limit for reactor coolant system pressure is 0 psia without uncertainty.

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Westinghouse Non-Proprietary Class 3 7-1 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E185-82

[Ref. 11] and is recommended for future capsules to be removed from the Salem Unit 2 reactor vessel. This recommended removal schedule is applicable to 50 EFPY of operation.

Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule Capsule Capsule Lead Withdrawal Fluence Location(a) Factor(a) EFPY(b) (n/cm , E > 1.0 MeV)(c) 2 T 40 3.41 1.19 2.75E+18 U 140 3.45 2.70 5.82E+18 X 220° 3.48 6.19 1.12E+19 Y 320° 3.47 10.80 1.82E+19 S 4° 1.38 (d) In reactor V 176° 1.38 (d) In reactor W 184° 1.38 (d) In reactor Z 356° 1.38 (d) In reactor Notes:

(a) Based on WCAP-15692 [Ref. 15].

(b) EFPY from plant startup.

(c) From Table 3-2.

(d) Capsule S, V, W, or Z could be designated as the fifth capsule to be withdrawn from the Salem Unit 2 reactor vessel to fulfill the requirements of ASTM E185-82, while the remaining three will then be standby capsules. The fifth capsule can be removed at any time after 35.4 EFPY but before the reactor vessel reaches 50 EFPY. It is recommended to be removed at 40 EFPY. At this time, the fifth capsule would have between one and two times the peak EOLR (50 EFPY) vessel fluence of 1.96E+19 n/cm2 (E > 1.0 MeV). It is recommended that the remaining three standby capsules also be removed and placed in storage at the time the fifth capsule is withdrawn as Section X1.M31 of NUREG-1801 [Ref. 3], Reactor Vessel Surveillance, states that any surveillance capsules that are left in the reactor vessel should provide meaningful metallurgical data. The NRC specifically states that anything beyond 60 years of exposure is not meaningful metallurgical data.

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Westinghouse Non-Proprietary Class 3 8-1 8 ERG P-T LIMITS CATEGORIZATION The ERG limits were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event [Ref. 19]. Generic categories of limits were developed for the guidelines based on the limiting inside surface RTNDT. These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants.

The highest value of RTNDT for which the generic category ERG limits were developed is 250°F for a longitudinal flaw and 300°F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250°F for a longitudinal flaw or 300°F for a circumferential flaw, plant-specific ERG P-T limits must be developed.

The ERG category is determined by the magnitude of the RTNDT value, which is equivalent to the RTPTS value as defined in Section (a)(7) of 10 CFR 50.61 [Ref. 14]. The material with the highest RTNDT defines the limiting material. The ERG limits are identified in Table 8-1.

Table 8-1 ERG P-T Limits Categories Applicable RTNDT Value(a) ERG P-T Limit Category RTNDT 200°F Category I 200°F < RTNDT 250°F Category II 250°F < RTNDT 300°F Category IIIb Note:

(a) Longitudinally oriented flaws are applicable only up to 250°F, the circumferentially oriented flaws are applicable up to 300°F.

Per the ERG limit guidance document [Ref. 19], some vessels do not change categories for operation through the end of license. However, when a vessel does change ERG categories between the beginning and end of operation, a plant-specific assessment must be performed to determine at what operating time the category changes. Thus, the ERG classification need not be changed until the operating cycle during which the maximum vessel value of actual or estimated real-time RTNDT exceeds the limit on its current ERG category. Note that for license extension, EOL in these discussions would be equal to EOLR.

The material with the highest RTNDT define the limiting material, which for Salem Unit 2 is the Lower Shell Longitudinal Weld Seams 3-442 A&C, with an EOLR RTPTS (RTNDT) value of 239°F (based on Table 4-1 of Section 4).

The operating time at which the ERG category would transition from Category I to Category II (i.e., when the RTNDT of the lower shell longitudinal weld seams 3-442 A&C would equal 200°F) was determined to be approximately 24.6 EFPY.

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8-2 Westinghouse Non-Proprietary Class 3 Conclusion of ERG P-T Limit Categorization The Salem Unit 2 reactor vessel material with the highest RTNDT value is the Lower Shell Longitudinal Weld Seams 3-442 A&C. Thus, this material is considered the limiting material for Salem Unit 2. The lower shell plate longitudinal weld seam 3-342 A&C has an EOLR RTPTS (RTNDT) value of 239°F (based on Table 4-1 of Section 4).

The operating time at which the ERG category would transition from Category I to Category II was determined to be approximately 24.6 EFPY.

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Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES

1. Code of Federal Regulations, 10 CFR Part 54.3, Definitions, August 28, 2007.

2. Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

3. NUREG-1801, Vol. 2, Rev. 1, Generic Aging Lessons Learned (GALL) Report, U. S.

Nuclear Regulatory Commission, Sept. 2005.

4. Fracture Toughness Requirements, Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

5. Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1988.

7. WCAP-15693, Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, April 2001.

8. PSEG Nuclear LLC Transmittal S-TODI-2008-0010, Design Clarifications for Upper Shelf Energy, RTNDT and ART, PTS, P-T Limits and Surveillance Capsule Withdrawal Schedule Evaluations and Analyses, September 22, 2008.

9. CMTR-RV-PNJ, PNJ Reactor Vessel Certified Materials Test Reports.

10. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Section NB-2300, Fracture Toughness Requirements for Material.

11. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels."

12. CE NPSD-1119, Revision 1, Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content, CEOG Task 1054, July 1998.

13. CEN-622-A Final Report, Generic Upper Shelf Values for Linde 1092, 124, and 0091 Reactor Vessel Welds, CEOG Task 839, December 1996.

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9-2 Westinghouse Non-Proprietary Class 3

14. Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", Federal Register, Volume 60, No.

143, dated December 19, 1995, effective January 18, 1996.

15. WCAP-15692, Revision 0, Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, August 2001.

16. Code of Federal Regulations, 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, February 20, 1971.

17. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure. [Sub-reference AMSE Section XI, Division 1, Code Case N-641, Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements, January 17, 2000.]

18. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

19. Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 2, April 30, 2005.

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIT)

The following pages contain the thermal stress intensity factors (KIT) for the maximum heatup and cooldown rates at 50 EFPY (Tables A-1 and A-2). The vessel radius to the 1/4T and 3/4T locations is as follows:

1/4T Radius = 88.875 inches

3/4T Radius = 93.188 inches WCAP-16982-NP December 2020 Revision 3

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A-2 Westinghouse Non-Proprietary Class 3 Table A-1 KIT Values for 100°F/hr. Heatup Curve for 50 EFPY Vessel Vessel Temperature Temperature @ 1/4T Thermal 3/4T Thermal Water @ 3/4T 1/4T Location Stress Stress Temp. Location for for 100°F/hr. Intensity Factor Intensity Factor

[°F] 100°F/hr.

Heatup [ksiin] [ksiin]

Heatup

[°F]

60 55.985 -0.9954 55.043 0.4731 65 58.558 -2.4522 55.294 1.4378 70 61.621 -3.7125 55.962 2.4257 75 64.898 -4.9101 57.099 3.3563 80 68.449 -5.9455 58.654 4.1903 85 72.111 -6.8918 60.589 4.9375 90 75.955 -7.7138 62.865 5.5993 95 79.898 -8.4650 65.439 6.1921 100 83.973 -9.1226 68.281 6.7187 105 88.134 -9.7208 71.356 7.1893 110 92.391 -10.2474 74.636 7.6094 115 96.721 -10.7282 78.096 7.9876 120 101.121 -11.1540 81.715 8.3272 125 105.582 -11.5440 85.475 8.6337 130 110.094 -11.8909 89.358 8.9099 135 114.655 -12.2101 93.350 9.1601 140 119.255 -12.4954 97.439 9.3868 145 123.894 -12.7592 101.614 9.5932 150 128.562 -12.9964 105.864 9.7812 155 133.262 -13.2171 110.180 9.9534 160 137.985 -13.4166 114.556 10.1111 165 142.732 -13.6036 118.983 10.2566 170 147.497 -13.7738 123.457 10.3907 175 152.281 -13.9343 127.971 10.5153 180 157.078 -14.0815 132.521 10.6310 185 161.892 -14.2214 137.104 10.7391 190 166.715 -14.3506 141.714 10.8403 195 171.551 -14.4744 146.349 10.9357 200 176.395 -14.5894 151.006 11.0256 205 181.248 -14.7005 155.683 11.1108 210 186.108 -14.8045 160.378 11.1918 WCAP-16982-NP December 2020 Revision 3

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Westinghouse Non-Proprietary Class 3 A-3 Table A-2 KIT Values for 100°F/hr. Cooldown Curve for 50 EFPY Vessel Temperature @ 100°F/hr Cooldown Water 1/4T Thermal 1/4T Location for Temp. Stress 100°F/hr.

[°F] Intensity Factor Cooldown

[ksiin]

[°F]

210 237.045 17.1391 205 231.959 17.0704 200 226.873 17.0021 195 221.787 16.9330 190 216.700 16.8642 185 211.613 16.7948 180 206.526 16.7258 175 201.439 16.6562 170 196.352 16.5870 165 191.265 16.5172 160 186.177 16.4480 155 181.089 16.3782 150 176.002 16.3089 145 170.914 16.2393 140 165.827 16.1700 135 160.739 16.1004 130 155.651 16.0313 125 150.564 15.9619 120 145.476 15.8929 115 140.389 15.8237 110 135.302 15.7549 105 130.214 15.6859 100 125.127 15.6173 95 120.040 15.5485 90 114.953 15.4802 85 109.866 15.4116 80 104.779 15.3435 75 99.693 15.2753 70 94.606 15.2075 65 89.520 15.1394 60 84.435 15.0711 WCAP-16982-NP December 2020 Revision 3

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LR-N22-0012 LAR S22-01 Enclosure 4 Westinghouse WCAP-18571-NP, Rev. 2, Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 WCAP-18571-NP July 2022 Revision 2 Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

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Westinghouse Non-Proprietary Class 3 WCAP-18571-NP Revision 2 Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation D. Brett Lynch*

RV/CV Design and Analysis Frank M. Nedwidek*

Nuclear Operations July 2022 Reviewer: Donald M. McNutt III*

RV/CV Design and Analysis Jared L. Geer*

Radiation Engineering & Analysis Approved: Lynn A. Patterson*, Manager RV/CV Design and Analysis Jesse J Klingensmith*, Manager Radiation Engineering & 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 ii RECORD OF REVISION Revision Description Completed 0 Original Issue September 2020 This revision corrects on error in Section 2.0 in 1 August 2021 identifying the cycles used for fluence projections.

This revision updates Reference 1, WCAP-16982-NP, to Revision 3. The change in referenced revision number 2 did not affect the content of this report. The Record of July 2022 Revision table was also added. Charges are marked with revision bars.

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iv LIST OF FIGURES ..................................................................................................................................... vi EXECUTIVE

SUMMARY

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

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 2-1 2.3 CALCULATIONAL UNCERTAINTIES ........................................................................ 2-3 3 FRACTURE TOUGHNESS PROPERTIES ................................................................................. 3-1 4 SURVEILLANCE DATA ............................................................................................................. 4-1 5 CHEMISTRY FACTORS ............................................................................................................. 5-1 6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .......................................... 6-1 7 P-T LIMIT CURVE APPLICABILITY EVALUATION .............................................................. 7-1 8 REFERENCES ............................................................................................................................. 8-1 APPENDIX A CREDIBILITY RE-EVALUATION OF THE SALEM UNIT 2 SURVEILLANCE PROGRAM ..................................................................................................................... A-1 APPENDIX B PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION ................................................................................................. B-1 APPENDIX C VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ................................................................................ C-1 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 2-1 RPV Material Locations .................................................................................................. 2-5 Table 2-2 Projection of Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Reactor Pressure Vessel Materials in the Beltline ......................................................................... 2-6 Table 2-3 Projection of Iron Atom Displacements (dpa) Experienced by Reactor Pressure Vessel Materials in the Beltline ................................................................................................... 2-7 Table 2-4 Surveillance Capsule Exposure Values ............................................................................ 2-8 Table 2-5 Projection of Surveillance Capsule Fluence and Lead Factors ........................................ 2-8 Table 2-6 Calculational Uncertainties .............................................................................................. 2-8 Table 3-1 Summary of the Best-Estimate Chemistry and Initial RTNDT Values for the Salem Unit 2 Reactor Vessel Materials .................................................................................................. 3-2 Table 3-2 Summary of the Best-Estimate Chemistry and Initial RTNDT Values for the Salem Unit 2 Reactor Vessel Inlet/Outlet Nozzle Materials .................................................................. 3-3 Table 4-1 Salem Unit 2 Surveillance Capsule Data ......................................................................... 4-2 Table 4-2 Sister-Plant Surveillance Program Results ...................................................................... 4-2 Table 5-1 Calculation of Chemistry Factors Using Salem Unit 2 Surveillance Capsule Data......... 5-2 Table 5-2 Calculation of Chemistry Factors Using Diablo Canyon Unit 2 Surveillance Capsule Data

......................................................................................................................................... 5-3 Table 5-3 Position 1.1 and 2.1 Chemistry Factors for Salem Unit 2................................................ 5-4 Table 6-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Salem Unit 2 Reactor Vessel Materials at 50 EFPY .................................................. 6-2 Table 6-2 Adjusted Reference Temperature Evaluation for the Salem Unit 2 Reactor Vessel Materials Through 50 EFPY at the 1/4T Location ........................................................... 6-3 Table 6-3 Adjusted Reference Temperature Evaluation for the Salem Unit 2 Reactor Vessel Materials Through 50 EFPY at the 3/4T Location ........................................................... 6-5 Table 6-4 Limiting ART Values for Salem Unit 2 at 50 EFPY ........................................................ 6-7 Table 7-1 Summary of the Limiting ART Values ............................................................................. 7-1 Table A-1 Identification of Credibility Criteria Affected by Fluence Change ................................. A-1 Table A-2 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Salem Unit 2 Surveillance Capsule Data Only .......................................................................... A-4 Table A-3 Surveillance Capsule Data Scatter about the Best-Fit Line Using Only Salem Unit 2 Surveillance Data ............................................................................................................ A-5 Table B-1 RTPTS Calculations for Salem Unit 2 Reactor Vessel Materials at 50 EFPY .................. B-2 Table B-2 Evaluation of Salem Unit 2 ERG Limit Category .......................................................... B-4 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 v Table C-1 Nuclear Parameters Used in the Evaluation of the In-Vessel Surveillance Capsule Neutron Sensors ............................................................................................................ C-10 Table C-2 Startup and Shutdown Dates ......................................................................................... C-10 Table C-3 Measured Sensor Activities and Reaction Rates for Surveillance Capsule T ............... C-11 Table C-4 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U .............. C-12 Table C-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X .............. C-13 Table C-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y .............. C-14 Table C-7 Comparison of Measured and Calculated Threshold Foil Reaction Rates for the In-Vessel Capsules ........................................................................................................................ C-15 Table C-8 Comparison of Calculated and Best-Estimate Exposure Rates for the In-Vessel Capsules

...................................................................................................................................... C-15 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 vi LIST OF FIGURES Figure 2-1 Plan View of the Reactor Geometry at the Core Midplane .............................................. 2-9 Figure 2-2 Section View of the Reactor Geometry - 45° Azimuth ................................................. 2-10 Figure 2-3 Section View of the Reactor Geometry - 4° Azimuth ................................................... 2-11 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 vii EXECUTIVE

SUMMARY

This report provides an evaluation of several parameters for reactor vessel (RV) integrity in order to determine whether the applicable U.S. Nuclear Regulatory Commission (NRC) requirements are met and continued safe operation of Salem Unit 2, with regards to RV integrity, can be justified through end-of-life extension (EOLE), i.e., 60 total years of operation. Specifically, this report documents the following Salem Unit 2 RV integrity calculations/evaluations:

1. Perform an updated neutron fluence assessment for the Salem Unit 2 pressure vessel materials.

Cycle-specific analyses for the past and current operating cycles, as well as projections for future operation through 50 effective full-power years (EFPY) are performed.

2. Determine the adjusted reference temperature (ART) values for the RV beltline and extended beltline materials at EOLE, 50 EFPY, i.e., 60 total years of operation for Unit 2.

3. Evaluate the applicability of the pressure-temperature (P-T) limit curves developed in WCAP-16982-NP [Ref. 1], with consideration of updated fluence values. The applicability of the Salem Unit 2 P-T limit curves remains unchanged at 50 EFPY.

4. Determine the pressurized thermal shock (PTS) reference temperature (RTPTS) values for the beltline and extended beltline materials in the RV at 50 EFPY. The RTPTS values of all of the beltline and extended beltline materials in the Salem Unit 2 RV are below the RTPTS screening criteria of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds (per 10 CFR 50.61.b.2), through EOLE (50 EFPY).

Appendix A contains a credibility evaluation for surveillance materials considering the updated fluence analysis.

Appendix B contains the evaluation for pressurized thermal shock (PTS) per 10 CFR 50.61 and emergency response guideline (ERG) analysis.

Appendix C contains a validation of the radiation transport calculation model based on neutron dosimetry measurements.

Revision 1 of this WCAP corrects on error in Section 2.0 in identifying the cycles used for fluence projections.

Revision 2 updates Reference 1, WCAP-16982-NP, to Revision 3.

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION The purpose of this report is to evaluate several parameters for reactor vessel (RV) integrity in order to determine whether the applicable U.S. Nuclear Regulatory Commission (NRC) requirements are met and continued safe operation of Salem Unit 2, with regards to RV integrity, can be justified through end-of-life extension (EOLE), i.e., 60 total years of operation.

As part of the time-limited aging analyses (TLAA) performed for the Salem Unit 2 license renewal, a fluence analysis was completed, and new heatup and cooldown P-T limit curves were developed, applicable to 50 EFPY. These analyses are documented in WCAP-16982-NP [Ref. 1].

In this report, a new fluence analysis is performed, using the fully three-dimensional RAPTOR-M3G methodology. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) are established on a plant- and fuel cycle-specific basis for the first 24 cycles.

Based on customer input, projections beyond the current cycle (Cycle 24) are based on Cycles 20 and 23.

All of the neutron transport calculations performed in this analysis are based on the nuclear cross-section data derived from ENDF/B-VI and make use of the latest available calculation tools. The neutron transport methodology follows the guidance of Regulatory Guide 1.190 [Ref. 2], and the methods used to determine the RV neutron exposures are consistent with the NRC approved methodology described in WCAP-18124-NP-A [Ref. 3].

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

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, the RTNDT increases associated with the new fluence projections discussed above need to be evaluated to ensure that the limiting RTNDT used in the 50 EFPY P-T limit curves in WCAP-16982-NP remain bounding. To find the most limiting RTNDT at any time period in the reactors life, RTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT. The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The NRC has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. 4].

Regulatory Guide 1.99, Revision 2 is used for the calculation of ART values (RTNDT(U) + RTNDT + margin for uncertainties) at the quarter thickness (1/4T) and three-quarter thickness (3/4T) locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface. These results are used to perform an evaluation to confirm the applicability of the P-T curves from WCAP-16982-NP with consideration of updated fluence values.

This report documents the calculated ART values in Section 6. A description of the updated fluence analysis is provided in Section 2 of this report, and a validation of the radiation transport calculation model based on neutron dosimetry measurements is contained in Appendix C.

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 1-2 Appendix A provides a credibility evaluation of the Salem Unit 2 surveillance data. Appendix B contains an evaluation of the Salem Unit 2 pressurized thermal shock (PTS) evaluation at EOLE and emergency response guideline (ERG) analysis.

WCAP-18571-NP July 2022 Revision 2

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

2.1 INTRODUCTION

Discrete ordinates (SN) transport analyses were performed to determine the neutron radiation environment within the reactor pressure vessel (RPV). In these analyses, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis. The dosimetry analysis documented in Appendix C shows that the

+/-20% (1) acceptance criteria specified in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [Ref. 2] is met, based on the measurement-to-calculation (M/C) comparison results for the in-vessel surveillance capsules withdrawn and analyzed to-date. These validated calculations form the basis for providing projections of the neutron exposure of the RPV through EOLE.

All of the calculations described in this section were based on nuclear cross-section data derived from the Evaluated Nuclear Data File (ENDF) database (specifically, ENDF/B-VI). Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET [Ref. 3]. The neutron transport evaluation methodology described in Reference 3 is based on the guidance of Regulatory Guide 1.190. Note, however, that the NRC Safety Evaluation Report (SER) in Reference 3 states that the applicability of the methodology described in Reference 3 is limited to the traditional RPV beltline region approximated by the RPV region near the active height of the core.

2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the RPV, a series of fuel-cycle-specific forward transport calculations were performed using the three-dimensional discrete ordinates code, RAPTOR-M3G

[Ref. 3], and the BUGLE-96 cross-section library [Ref. 5]. The BUGLE-96 library provides a coupled 47-neutron and 20-gamma-ray group cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modeled with an S8 order of angular quadrature. Energy- and space-dependent core power distributions were treated on a fuel-cycle-specific basis.

The Salem Unit 2 reactor is a standard Westinghouse 4-loop design employing reactor internals that include 1.125-inch-thick baffle plates and a fully circumferential thermal shield. The model of the reactor (and reactor cavity) geometry used in the plant-specific evaluation is shown in Figure 2-1 through Figure 2-3.

The model extends radially from the center of the core to 350 cm, azimuthally from 0° to 45° (taking advantage of the octant symmetry of the reactor configuration), and axially from -325 cm to 325 cm with respect to the midplane of the active core. Elevations of key RPV materials relative to the model geometry are provided in Table 2-1.

A plan view of the model geometry at the core midplane is shown in Figure 2-1. In this figure, a single octant is depicted showing the arrangement of the core, reactor internals, core barrel, thermal shield, downcomer, cladding, RPV, reactor cavity, reflective insulation, and bioshield. Depictions of the in-vessel surveillance capsules, including their associated support structures, are also shown.

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-2 From a neutronics standpoint, the inclusion of the surveillance capsules and associated support structures in the geometric model is significant. Since the presence of the capsules and support structures has a marked impact on the magnitude of the neutron fluence rate and relative neutron and gamma ray spectra at dosimetry locations within the capsules, a meaningful evaluation of the radiation environment internal to the capsules can be made only when these perturbation effects are accounted for in the transport calculations.

A section view of the model geometry is shown in Figure 2-2 and Figure 2-3. Note that the stainless-steel former plates located between the core baffle and barrel regions are shown in these figures.

When developing the reactor model shown in Figure 2-1 through Figure 2-3, nominal design dimensions were employed for the various structural components. Likewise, water temperatures and, hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. These coolant temperatures were varied on a cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids and guide tubes.

The geometric mesh description of the reactor model shown in Figure 2-1 through Figure 2-3 consisted of 126 radial by 68 azimuthal by 195 axial intervals. Mesh sizes were chosen to ensure sufficient resolution of the stair-step shaped baffle plates as well as an adequate number of meshes throughout the radial and axial regions of interest. The pointwise inner iteration convergence criterion utilized in the calculations was set at a value of 0.001.

The core power distributions used in the plant-specific transport analysis were taken from nuclear design documentation. The data extracted included fuel assembly-specific initial enrichments, beginning-of-cycle burnups and end-of-cycle burnups. Appropriate axial power distributions were also obtained.

For each fuel cycle of operation, fuel-assembly-specific enrichment and burnup data were used to generate the spatially dependent neutron source throughout the reactor core. This source description included the spatial variation of isotope-dependent (U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242) fission spectra, neutron emission rate per fission, and energy release per fission based on the burnup history of individual fuel assemblies. These fuel-assembly-specific neutron source strengths derived from the detailed isotopics were then converted from fuel pin Cartesian coordinates to the spatial mesh arrays used in the discrete ordinates calculations.

In Table 2-1, axial and azimuthal locations of the RPV materials are provided. The axial position of each material is indexed to z = 0.0 cm, which corresponds to the midplane of the active fuel stack.

Cycle-specific calculations were performed for Cycles 1-24. Note that future fluence projection data beyond Cycle 24 are based on Cycles 20 and 23. At the time of the development of the fluence model, Cycle 24 had not been completed and, thus, the results for Cycle 24 are based on the cycle design data.

Neutron fluence projections for the RPV are given in Table 2-2. Similarly, iron atom displacement results for the RPV are provided in Table 2-3. The data presented represent the maximum neutron exposures experienced by RPV materials. These projections were based on the spatial power distribution and reactor operating conditions of Cycles 20 and 23. The projected results will remain valid as long as future plant WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-3 operation is consistent with these assumptions. Results of the discrete ordinates transport analyses pertinent to the surveillance capsule evaluations are provided in Appendix C. The reported data also consider both the inner and outer radius of the RPV base metal, and account for the possibility of higher neutron exposure values occurring on the outer surface of the RPV (as compared to the inner surface) for materials that are distant from the active core.

To allow for the determination of potential fast fluence accumulation, the projected fast fluence rate (E > 1.0 MeV) at each surveillance capsule location is provided in Table 2-5. Projections of future operation are based on the spatial power distributions and reactor operating conditions of Cycles 20 and 23. Note that RPV neutron exposure rates are dominated by neutron leakage from the peripheral fuel assemblies. The additional fast fluence accumulated for any re-inserted/re-located capsule can be determined by multiplying the fast fluence rate value in Table 2-5 for the appropriate capsule position with the irradiation duration in effective full-power seconds (EFPS).

2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the RPV is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology used in the plant-specific neutron exposure evaluation is carried out in the following four stages:

1. Comparisons of calculations with benchmark measurements from the pool critical assembly (PCA) simulator (NUREG/CR-6454, Pool Critical Assembly Pressure Vessel Facility Benchmark [Ref. 6]) at the Oak Ridge National Laboratory (ORNL) and the VENUS-1 experiment.

2. Comparison of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment (NUREG/CR-6453, H.B.

Robinson-2 Pressure Vessel Benchmark [Ref. 7]).

3. An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments (WCAP-18124-NP-A [Ref. 3]).

4. Comparison of the calculations with all available dosimetry results from the RPV measurement programs carried out at Salem Unit 2 (Appendix C).

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations, nor did it address uncertainties in operational and geometric variables that impact power reactor calculations.

The second phase of the qualification (H.B. Robinson comparisons) addressed uncertainties that are primarily methods-related and would tend to apply generically to all fast neutron exposure evaluations.

The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational method approximations as well as to a lack of knowledge relative to various plant-specific parameters. The overall calculational uncertainty applicable to the Salem Unit 2 analyses were established from the results of these three phases of the methods qualification.

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-4 The fourth phase of the uncertainty assessment (comparisons of plant-specific dosimetry measurements) was used solely to demonstrate the adequacy of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used to bias the final results in any way.

Table 2-6 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in WCAP-18124-NP-A. The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Table 2-4 support these uncertainty assessments for Salem Unit 2.

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-5 Table 2-1 RPV Material Locations Azimuthal Location(a) Axial Location(b)

Material [°] [cm]

Min Max Min Max Outlet nozzle to nozzle shell welds

- Nozzle 1 22.0 270.12

- Nozzle 2 158.0 270.12

- Nozzle 3 202.0 270.12

- Nozzle 4 338.0 270.12 Inlet nozzle to nozzle shell welds

- Nozzle 1 67.0 265.04

- Nozzle 2 113.0 265.04

- Nozzle 3 247.0 265.04

- Nozzle 4 293.0 265.04 Nozzle shell

- Plate 1 300.0 60.0 233.53 485.07

- Plate 2 60.0 180.0 233.53 485.07

- Plate 3 180.0 300.0 233.53 485.07 Nozzle shell longitudinal welds

- Weld 1 60.0 233.53 485.07

- Weld 2 180.0 233.53 485.07

- Weld 3 300.0 233.53 485.07 Nozzle shell to intermediate shell circumferential weld

- centerline 0.0 360.0 233.53 Intermediate shell

- Plate 1 0.0 120.0 -42.93 233.53

- Plate 2 120.0 240.0 -42.93 233.53

- Plate 3 240.0 360.0 -42.93 233.53 Intermediate shell longitudinal welds

- Weld 1 0.0 -42.93 233.53

- Weld 2 120.0 -42.93 233.53

- Weld 3 240.0 -42.93 233.53 Intermediate shell to lower shell circumferential weld

- centerline 0.0 360.0 -42.93 Lower shell

- Plate 1 300.0 60.0 -313.52 -42.93

- Plate 2 60.0 180.0 -313.52 -42.93

- Plate 3 180.0 300.0 -313.52 -42.93 Lower shell longitudinal welds

- Weld 1 60.0 -313.52 -42.93

- Weld 2 180.0 -313.52 -42.93

- Weld 3 300.0 -313.52 -42.93 Lower shell to lower vessel head circumferential weld

- centerline 0.0 360.0 -313.52 Notes:

(a) Azimuthal locations are indexed to = 0.0.

(b) Axial elevations are indexed to Z = 0.0 at the midplane of the active fuel stack.

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-6 Table 2-2 Projection of Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Reactor Pressure Vessel Materials in the Beltline Neutron Fluence (n/cm2)

Beltline Material 32 EFPY 48 EFPY 50 EFPY Vessel shells Lower shell 1.34E+19 1.93E+19 2.01E+19 Intermediate shell 1.36E+19 1.97E+19 2.05E+19 Upper shell 1.71E+17 2.36E+17 2.44E+17 Circumferential (or girth) welds Bottom torus/lower shell 2.21E+15 3.12E+15 3.24E+15 Lower shell/intermediate shell 1.32E+19 1.91E+19 1.98E+19 Intermediate shell/upper shell 1.96E+17 2.71E+17 2.80E+17 Longitudinal welds Lower shell 60° and 300° 9.56E+18 1.39E+19 1.45E+19 180° 5.00E+18 7.11E+18 7.38E+18 Intermediate shell 0° 4.97E+18 7.14E+18 7.41E+18 120° and 240° 9.61E+18 1.41E+19 1.47E+19 Upper shell 60° and 300° 1.27E+17 1.77E+17 1.83E+17 180° 7.10E+16 9.77E+16 1.01E+17 Nozzle forging attachment weld Inlet nozzle 1.56E+16 2.18E+16 2.26E+16 Outlet nozzle 1.23E+16 1.72E+16 1.78E+16 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-7 Table 2-3 Projection of Iron Atom Displacements (dpa)

Experienced by Reactor Pressure Vessel Materials in the Beltline Iron Atom Displacements (dpa)

Beltline Material 32 EFPY 48 EFPY 50 EFPY Vessel shells Lower shell 2.16E-02 3.11E-02 3.23E-02 Intermediate shell 2.20E-02 3.19E-02 3.31E-02 Upper shell 3.10E-04 4.29E-04 4.44E-04 Circumferential (or girth) welds Bottom torus/lower shell 1.40E-05 1.97E-05 2.04E-05 Lower shell/intermediate shell 2.14E-02 3.10E-02 3.21E-02 Intermediate shell/upper shell 3.55E-04 4.90E-04 5.07E-04 Longitudinal welds Lower shell 60° and 300° 1.54E-02 2.24E-02 2.33E-02 180° 8.02E-03 1.14E-02 1.19E-02 Intermediate shell 0° 8.06E-03 1.16E-02 1.20E-02 120° and 240° 1.55E-02 2.28E-02 2.37E-02 Upper shell 60° and 300° 2.34E-04 3.25E-04 3.37E-04 180° 1.31E-04 1.81E-04 1.87E-04 Nozzle forging attachment weld Inlet nozzle 4.15E-05 5.89E-05 6.10E-05 Outlet nozzle 3.44E-05 4.88E-05 5.06E-05 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-8 Table 2-4 Surveillance Capsule Exposure Values Capsule Capsule Location Fluence (First-Octant- Withdrawn Withdrawal (E > 1.0 MeV)

Capsule Equivalent) (EOC) (EFPY) (n/cm2) Lead Factor 40° T 1 1.2 2.73E+18 3.31 (40°)

140° U 3 2.7 5.81E+18 3.35 (40°)

220° X 6 6.2 1.13E+19 3.36 (40°)

320° Y 11 10.8 1.83E+19 3.36 (40°)

Table 2-5 Projection of Surveillance Capsule Fluence and Lead Factors Capsule Fluence (E > 1.0 MeV)

(n/cm2) Lead Factor 4° 40° 4° 40° EFPY Capsule Capsule Capsule Capsule 32 1.57E+19 4.57E+19 1.15 3.36 48 2.25E+19 6.62E+19 1.14 3.36 50 2.34E+19 6.88E+19 1.14 3.36 Table 2-6 Calculational Uncertainties Uncertainty Description Vessel Inner Capsule Radius PCA comparisons 3% 3%

H.B. Robinson comparisons 5% 5%

Analytical sensitivity studies 9% 11%

Additional uncertainty for factors not explicitly evaluated 5% 5%

Net calculational uncertainty 12% 13%

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Westinghouse Non-Proprietary Class 3 2-9 Figure 2-1 Plan View of the Reactor Geometry at the Core Midplane WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-10 Figure 2-2 Section View of the Reactor Geometry - 45° Azimuth WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 2-11 Figure 2-3 Section View of the Reactor Geometry - 4° Azimuth WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The requirements for P-T limit curve development are specified in 10 CFR 50, Appendix G [Ref. 8]. The beltline region of the reactor vessel is defined as the following in 10 CFR 50, Appendix G:

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

The Salem Unit 2 beltline materials traditionally included the intermediate and lower shell plates and welds; however, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. 9], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period should be considered in the development of P-T limit curves. The additional materials that exceed this fluence threshold are referred to as the extended beltline materials and are evaluated to ensure that the applicable neutron embrittlement effects are considered. As seen from Table 2-2 of this report, the extended beltline materials include upper shell plates, upper shell longitudinal welds, and the upper to intermediate shell girth weld. The fluence for both the inlet and outlet nozzle to upper shell welds are less than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at 50 EFPY. Therefore, the materials of the inlet/outlet nozzle forgings and the associated welds to the upper shell do not need to be considered in the extended beltline. Note that for reactor vessel welds, the terms girth and circumferential are used interchangeably; herein, these welds shall be referred to as circumferential welds. Similarly, for reactor vessel welds, the terms axial and longitudinal are used interchangeably; herein, these welds shall be referred to as longitudinal welds.

Although the reactor vessel nozzles are not a part of the extended beltline, per NRC RIS 2014-11, the nozzle materials must be evaluated for their potential effect on P-T limit curves due to the higher stresses in the nozzle corner region. These higher stresses can potentially result in more restrictive P-T limits, even if the RTNDT for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries. The effect of these higher stresses is addressed in Section 7, which determines that the Salem Unit 2 beltline P-T limit curves generated in WCAP-16982-NP [Ref. 1] bound the inlet and outlet nozzle P -T limit curves.

A summary of the best-estimate copper (Cu) and nickel (Ni) contents in units of weight percent (wt. %), as well as initial RTNDT values, for the Salem Unit 2 reactor vessel beltline and extended beltline materials are provided in Table 3-1. Table 3-2 provides the best-estimate Cu and Ni wt. % values and initial RTNDT values for the reactor vessel nozzle forgings.

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Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of the Best-Estimate Chemistry and Initial RTNDT Values for the Salem Unit 2 Reactor Vessel Materials(a)

Heat Wt. % Wt. % RTNDT(U) I(b)

Material Description Flux Type (Lot)

Number Cu Ni (°F) (°F)

Upper Shell B4711-1 C4194-1 --- 0.11 0.55 60 0 Upper Shell B4711-2 C4149-2 --- 0.14 0.56 60 0 Upper Shell B4711-3 C4171-1 --- 0.12 0.58 101 0 Intermediate Shell B4712-1 C4173-1 --- 0.13 0.56 0 0 Intermediate Shell B4712-2 C4186-2 --- 0.12 0.61 12 0 Intermediate Shell B4712-3 C4194-2 --- 0.11 0.57 10 0 Lower Shell B4713-1 C4182-1 --- 0.12 0.60 8 0 Lower Shell B4713-2 C4182-2 --- 0.12 0.57 8 0 Lower Shell B4713-3 B-8343-1 --- 0.12 0.58 10 0 Intermediate to Upper Shell Linde 1092 20291/1P2809 0.27(c) 0.735(c) -56(d) 17(d)

Circumferential Weld Seam 8-442 (Lot # 3854)

Intermediate to Lower Shell Linde 0091 90099 0.197 0.06 -56(d) 17(d)

Circumferential Weld Seam 9-442 (Lot # 3977)

Upper Shell Longitudinal Weld Linde 1092 13253/12008 0.21(c) 0.873(c) -56(d) 17(d)

Seams 1-442 A, B, & C (Lot # 3833)

Intermediate Shell Longitudinal Linde 1092 13253/20291 0.221 0.732 -56(d) 17(d)

Weld Seams 2-442 A, B, & C (Lot # 3833)

Lower Shell Longitudinal Weld Linde 1092 21935/12008 0.213 0.867 -56(d) 17(d)

Seams 3-442 A, B, & C (Lot # 3889)

Salem Unit 2 Surveillance Weld Linde 1092 13253 0.225 0.727 --- ---

Material(f) (Lot # 3883/3774)

Diablo Canyon Unit 2 Surveillance Linde 1092 21935/12008(e) 0.22(e) 0.87(e) --- ---

Program Weld Metal (Lot # 3869)(e)

Notes:

(a) The data was extracted from WCAP-16982-NP [Ref. 1], unless otherwise noted.

(b) All RTNDT(U) values are based on measured data with a I = 0°F, unless otherwise noted.

(c) Best estimate copper and nickel content based on CE NPSD-1119, Revision 1 [Ref. 10]. Note that in the case of the Intermediate to Upper Shell Circumferential Weld Seam 8-442, there is no record that the two Heat Numbers 20291 and 1P2809 were deposited in tandem; therefore, the copper and nickel content for Heat # 1P2809 is used as a more conservative basis for estimating the effect of neutron irradiation.

(d) This value is a generic weld RTNDT(U) value from 10 CFR 50.61 [Ref. 11]. The use of this estimated RTNDT(U) requires I = 17°F.

(e) The Diablo Canyon Unit 2 surveillance information is taken from WCAP-15423 [Ref. 12].

(f) Note that the Salem Unit 2 surveillance weld metal Heat # 13253 is only representative of the beltline welds, not identical. Therefore, the surveillance weld data was not used in the calculations documented in this report.

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Westinghouse Non-Proprietary Class 3 3-3 Table 3-2 Summary of the Best-Estimate Chemistry and Initial RTNDT Values for the Salem Unit 2 Reactor Vessel Inlet/Outlet Nozzle Materials(a)

Heat Flux Type Wt. % Wt. % RTNDT(U) I(b)

Material Description Number (Lot) Cu Ni (°F) (°F)

Inlet Nozzle B4703-1 ZV-3265 --- N/A 0.69 -27 0 Inlet Nozzle B4703-2 ZV-3265 --- N/A 0.685 -29 0 Inlet Nozzle B4703-3 ZV-3265 --- N/A 0.675 -23 0 Inlet Nozzle B4703-4 SV 2040-1 --- N/A 0.805 -31 0 Outlet Nozzle B4704-1 AV-2042 --- N/A 0.835 -68 0 Outlet Nozzle B4704-2 AV-2061 --- N/A 0.77 -44 0 Outlet Nozzle B4704-3 AV-2067 --- N/A 0.69 -72 0 Outlet Nozzle B4704-4 AV-2099 --- N/A 0.71 -51 0 Notes:

(a) The inlet/outlet nozzle forgings are projected to be less than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at EOLE and do not need to be considered for the effects of embrittlement. However, the initial material properties are provided here for future reference. Note the initial RTNDT values were calculated, consistent with the Unit 1 nozzles in WCAP-18502-NP [Ref. 13], using the methodology in BWRVIP-173-A [Ref. 14] to supplement the ASME Section III methodology to account for limited material testing data available.

(b) All RTNDT(U) values are based on measured data with I = 0°F.

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Westinghouse Non-Proprietary Class 3 4-1 4 SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2 [Ref. 4], calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, data from surveillance programs at other plants, which include a Salem Unit 2 reactor vessel beltline or extended beltline materials, should also be considered when calculating Position 2.1 chemistry factors. Data from a surveillance program at another plant is often called sister-plant data.

The Salem Unit 2 surveillance capsules contain plate material from Intermediate Shell Plate B4712-2.

Table 4-1 summarizes the surveillance data available for the Salem Unit 2 plate materials that will be used in the calculation of the Position 2.1 chemistry factor values. Per Appendix A, the surveillance data are deemed credible for Salem Unit 2.

The Salem Unit 2 surveillance weld specimens were fabricated from weld wire Heat # 13253. Table 4-1 summarizes the surveillance data available for the Salem Unit 2 weld materials. However, the Salem Unit 2 surveillance weld wire Heat # 13253 is only representative of the beltline welds, not identical. Therefore, the surveillance weld data are not used, and the data are provided in Table 4-1 for information only.

The Diablo Canyon Unit 2 surveillance program contains weld wire Heat # 21935/12008 with Linde 1092 flux, which was also used in the fabrication of Salem Unit 2 Lower Shell Longitudinal Weld Seams 3-442 A, B, & C. Thus, the data from this surveillance program is applicable to Salem Unit 2. Table 4-2 contains the sister-plant weld material surveillance data. Per WCAP-17315-NP [Ref. 15], the surveillance data are deemed credible; therefore, a reduced margin term will be utilized in the ART calculations contained in Section 6 for the Salem Unit 2 lower shell longitudinal weld seams.

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Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 Salem Unit 2 Surveillance Capsule Data(a)

Capsule Measured Irradiation Fluence(b)

Material Capsule RTNDT Temperature (x 1019 n/cm2,

(°F) (°F)

E > 1.0 MeV)

T 0.273 61.66 545 Intermediate Shell Plate B4712-2 U 0.581 66.54 543 (longitudinal) X 1.13 93.82 541 Y 1.83 105.69 540 T 0.273 74.83 545 Intermediate Shell Plate B4712-2 U 0.581 98.26 543 (transverse) X 1.13 125.15 541 Y 1.83 129.33 540 T 0.273 153.17 545 Salem Unit 2 Surveillance Weld U 0.581 185.94 543 Material(c)

(Heat #13253) X 1.13 195.43 541 Y 1.83 200.90 540 Notes:

(a) Information extracted from WCAP-15692 [Ref. 16], unless otherwise noted.

(b) The fluence values are taken from Table 2-4.

(c) Note that the Salem Unit 2 surveillance weld metal Heat # 13253 is only representative of the beltline welds, not identical. Therefore, the surveillance weld daata was not used in the calculations documented in this report.

Table 4-2 Sister-Plant Surveillance Program Results Withdrawal Capsule Fluence Measured Irradiation Material Capsule (x 1019 n/cm2, RTNDT Temperature Cycle EFPY E > 1.0 MeV) (°F) (°F)

U 1 1.02 0.330 173.0 545 Diablo Canyon Unit 2 Surveillance Weld X 3 3.16 0.906 203.2 545 Material(a) Y 6 7.08 1.53 211.4 545 (Heat # 21935/12008)

V 9 11.49 2.38 224.5 545 Note:

(a) The Diablo Canyon Unit 2 capsule data were taken from WCAP-17315-NP [Ref. 15]. EFPY values are found in WCAP-15423 [Ref. 12].

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Westinghouse Non-Proprietary Class 3 5-1 5 CHEMISTRY FACTORS The chemistry factors (CFs) are calculated using Regulatory Guide 1.99, Revision 2 [Ref. 4], Positions 1.1 and 2.1. Position 1.1 chemistry factors for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.

The best-estimate copper and nickel weight percent values for the Salem Unit 2 reactor vessel materials are provided in Table 3-1 of this report. Regulatory Guide 1.99, Position 1.1 chemistry factors are listed in Table 5-3.

The Position 2.1 CFs are calculated for the materials that have available surveillance data from the plant-specific surveillance program. The Position 2.1 CF calculation is performed using the method described in Regulatory Guide 1.99, Revision 2. The Salem Unit 2 surveillance data are summarized in Section 4 of this report and will be utilized in the Position 2.1 CF calculations in this section. The Position 2.1 CF calculations are presented in Table 5-1 for the Salem Unit 2 surveillance materials. The Position 2.1 CF calculations are presented in Table 5-2 for the sister-plant surveillance material.

In addition to the plant-specific surveillance data, data from surveillance programs at other plants which include a Salem Unit 2 reactor vessel beltline or extended beltline material should also be considered when calculating Position 2.1 CFs. As discussed in Section 4, Salem Unit 2 does utilize surveillance data from a sister plant in the beltline or extended beltline. Adjustment of the RTNDT values are required per Regulatory Guide 1.99, Revision 2 due to chemistry differences between the surveillance welds and the Salem Unit 2 reactor vessel welds. The chemistry adjustment factor based on the differences between the Regulatory Guide 1.99, Position 1.1 CFs are determined below.

Salem Unit 2 Lower Shell Longitudinal Weld Seams 3-442A, 3-442B, and 3-442C (Heat # 21935/12008)

Diablo Canyon Unit 2 data CFBeltline Weld (Salem 2) = 208.6°F CFSurv. Weld (Diablo Canyon 2) = 211.2°F Ratio = 208.6 / 211.2 = 0.99 The ratio procedure results in a ratio of less than 1.0; therefore, the chemistry adjustment can conservatively be neglected.

From NRC Industry Meetings on November 12, 1997 and February 12 and 14, 1998, procedural guidelines were presented to adjust the RTNDT for temperature differences when using surveillance data from one reactor vessel applied to another reactor vessel. The following is taken from the handout [Ref. 17] given by the NRC at those industry meetings.

Studies have shown that for temperatures near 550°F, a 1°F decrease in irradiation temperature will result in approximately a 1°F increase in RTNDT.

Thus, for plants that use surveillance data from other reactor vessels that operate at a different temperature, or when the capsule is at a different temperature than the plant, then this difference must be considered. The temperature adjustment procedure is not applied when considering only surveillance data from the plant-specific program. The temperature adjustment is as follows:

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Westinghouse Non-Proprietary Class 3 5-2 Temp. Adjusted RTNDT = RTNDT, Measured + (Tcapsule - TPlant) (1)

When the surveillance data from capsules irradiated in other plants are applied to the Salem Unit 2 reactor vessel welds, the data must be adjusted to the Salem Unit 2 temperature, Tcold, using the equation listed above. The time-averaged irradiation temperature of the Salem Unit 2 vessel over the lifetime of the plant is 542°F. The irradiated temperatures of the surveillance capsules from Diablo Canyon Unit 2 are provided in Table 4-2.

The Position 1.1 and Position 2.1 chemistry factors are summarized in Table 5-3 for Salem Unit 2.

Table 5-1 Calculation of Chemistry Factors Using Salem Unit 2 Surveillance Capsule Data(a)

Capsule Fluence RTNDT FF* RTNDT Material Capsule FF(b) FF2 (x 1019 n/cm2, (°F) (°F)

E > 1.0 MeV)

T 0.273 0.646 61.66 39.84 0.418 Intermediate Shell Plate U 0.581 0.848 66.54 56.43 0.719 B4712-2 X 1.13 1.034 93.82 97.02 1.069 (longitudinal)

Y 1.83 1.166 105.69 123.21 1.359 T 0.273 0.646 74.83 48.35 0.418 Intermediate Shell Plate U 0.581 0.848 98.26 83.33 0.719 B4712-2 X 1.13 1.034 125.15 129.42 1.069 (transverse)

Y 1.83 1.166 129.33 150.76 1.359 SUM: 728.36 7.130 CFB4712-2 = (FF

RTNDT) ÷ (FF2) = (728.36) ÷ (7.130) = 102.2°F T 0.273 0.646 153.17 98.97 0.418 U 0.581 0.848 185.94 157.68 0.719 Salem Unit 2 Surveillance X 1.13 1.034 195.43 202.10 1.069 Weld Material Y 1.83 1.166 200.90 234.20 1.359 (Heat # 13253)

SUM: 692.95 3.565 2

CFSurv. Weld = (FF

RTNDT) ÷ (FF ) = (692.95) ÷ (3.565) = 194.4°F Notes:

(a) Fluence and measured RTNDT are taken from Table 4-1.

(b) FF = fluence factor = f(0.28 - 0.10l*log(f)).

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Westinghouse Non-Proprietary Class 3 5-3 Table 5-2 Calculation of Chemistry Factors Using Diablo Canyon Unit 2 Surveillance Capsule Data(a)

Capsule Adjusted Fluence FF* RTNDT Material Capsule FF (b)

RTNDT(c) FF2 (x 1019 n/cm2, (°F)

(°F)

E > 1.0 MeV) 176.0 U 0.330 0.695 122.32 0.483 (173.0) 206.2 X 0.906 0.972 200.49 0.945 (203.2)

Diablo Canyon Unit 2 Surveillance 214.4 Y 1.53 1.118 239.62 1.249 Weld Material (211.4)

(Heat # 21935/12008) 227.5 V 2.38 1.234 280.70 1.522 (224.5)

SUM: 843.14 4.200 CFSurv. Weld = (FF

RTNDT) ÷ (FF2) = (843.14) ÷ (4.200) = 200.7°F Notes:

(a) Fluence and measured RTNDT taken from Table 4-2.

(b) FF = fluence factor = f(0.28 - 0.10l*log(f)).

(c) The surveillance weld RTNDT values have been adjusted for irradiation temperature as follows:

Adjusted RTNDT = RTNDT, Measured + Temp. Adjustment The temperature adjustments are based on a Salem Unit 2 reactor vessel temperature of 542°F. The Diablo Canyon Unit 2 capsule irradiation temperatures are provided in Table 4-2. The measured (unadjusted) RTNDT values are shown in parenthesis. The temp. adjustment is therefore 3°F (545°F - 542°F).

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Westinghouse Non-Proprietary Class 3 5-4 Table 5-3 Position 1.1 and 2.1 Chemistry Factors for Salem Unit 2 Chemistry Factor Material Position 1.1(a) Position 2.1(b)

(°F) (°F)

Upper Shell B4711-1 73.5 ---

Upper Shell B4711-2 98.2 ---

Upper Shell B4711-3 82.6 ---

Intermediate Shell B4712-1 89.8 ---

Intermediate Shell B4712-2 83.2 102.2 Intermediate Shell B4712-3 73.7 ---

Lower Shell B4713-1 83.0 ---

Lower Shell B4713-2 82.4 ---

Lower Shell B4713-3 82.6 ---

Intermediate to Upper Shell 205.6 ---

Circumferential Weld Seam 8-442 Intermediate to Lower Shell 91.4 ---

Circumferential Weld Seam 9-442 Upper Shell Longitudinal Weld Seams 208.7 ---

1-442 A, B, & C Intermediate Shell Longitudinal Weld 189.1 ---

Seams 2-442 A, B, & C Lower Shell Longitudinal Weld Seams 208.6 200.7 3-442 A, B, & C Salem Unit 2 Surveillance Weld 189.3 ---

Material Diablo Canyon Unit 2 Surveillance 211.2 ---

Program Weld Metal Notes:

(a) All values are basd on Tables 1 and 2 of Regulatory Guide 1.99, Revision 2 [Ref. 4] (Position 1.1) and the Cu and Ni weight percent values given in Table 3-1.

(b) Values are from Table 5-1 or Table 5-2.

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Westinghouse Non-Proprietary Class 3 6-1 6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2 [Ref. 4], the ART for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + RTNDT + Margin (2)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [Ref. 18]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided there are sufficient test results to establish a mean and standard deviation for the class.

RTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

RTNDT = CF

f(0.28 - 0.10 log (f)) (3)

To calculate RTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth:

f(depth x) = fsurf

e(-0.24x) (4)

Where:

x = depth into the vessel wall measured from the vessel clad/base metal interface in inches (reactor vessel cylindrical shell beltline thickness is 8.625 inches).

The resultant fluence is then placed in Equation 3 to calculate the RTNDT at the specific depth. The projected reactor vessel neutron fluence was updated for this analysis and is documented in Section 2 of this report.

Table 6-1 contains the surface fluence values at 50 EFPY, as well as the 1/4T and 3/4T calculated fluence values and fluence factors (FFs), per Regulatory Guide 1.99, Revision 2. The values in this table will be used to calculate the EOLE ART values for the Salem Unit 2 reactor vessel materials.

Margin is calculated as = 2 2 + 2. The standard deviation for the initial RTNDT margin term (I) is 0°F when the initial RTNDT is a measured value. When a generic value is used, the I is obtained from the set of data used to establish the mean. The standard deviation for the RTNDT margin term () is 17°F for plates or forgings when surveillance data are not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data are used. For welds, is equal to 28°F when surveillance capsule data are not used or is non-credible and is 14°F (half the value) when credible surveillance capsule data are used. The value for need not exceed 0.5 times the mean value of RTNDT.

Per Appendix A, the surveillance data for Intermediate Shell Plate B4712-2 are deemed credible. The surveillance weld data for Heat # 21935/12008 from the Diablo Canyon Unit 2 surveillance program are also deemed credible.

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Westinghouse Non-Proprietary Class 3 6-2 Tables 6-2 and 6-3 contain the 50 EFPY ART calculations at the 1/4T and 3/4T locations, respectively. The limiting ART values are summarized in Table 6-4.

Table 6-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Salem Unit 2 Reactor Vessel Materials at 50 EFPY Surface 1/4T 3/4T Fluence(a) Surface Fluence(b) 1/4T Fluence(b) 3/4T Material (x 1019 n/cm2, FF(c) (x 1019 n/cm2, FF(c) (x 1019 n/cm2, FF(c)

E > 1.0 MeV) E > 1.0 MeV) E > 1.0 MeV)

Upper shell plates 0.0244 0.194 0.0145 0.141 0.00517 0.069 Intermediate shell plates 2.05 1.196 1.22 1.056 0.434 0.768 Lower shell plates 2.01 1.190 1.20 1.050 0.426 0.763 Intermediate to Upper Shell Circumferential Weld Seam 0.028 0.211 0.0167 0.154 0.00593 0.076 8-442 Intermediate to Lower Shell Circumferential Weld Seam 1.98 1.186 1.18 1.046 0.419 0.759 9-442 Upper Shell Longitudinal 0.0183 0.163 0.0109 0.116 0.00387 0.055 Weld Seams 1-442 A & C Upper Shell Longitudinal 0.0101 0.110 0.0060 0.077 0.00214 0.035 Weld Seam 1-442 B Intermediate Shell Longitudinal Weld Seams 1.47 1.107 0.876 0.963 0.311 0.680 2-442 B & C Intermediate Shell Longitudinal Weld Seam 0.741 0.916 0.442 0.773 0.157 0.513 2-442 A Lower Shell Longitudinal 1.45 1.103 0.864 0.959 0.307 0.676 Weld Seams 3-442 A & C Lower Shell Longitudinal 0.738 0.915 0.440 0.772 0.156 0.512 Weld Seam 3-442 B Notes:

(a) Fluence values are documented in Table 2-2.

(b) The 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (8.625 inches), and equation f = fsurf*e-0.24(x) from Regulatory Guide 1.99, Revision 2, where x = depth into the vessel wall (inches).

(c) FF = fluence factor = f(0.28 - 0.10 log(f)).

WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 6-3 Table 6-2 Adjusted Reference Temperature Evaluation for the Salem Unit 2 Reactor Vessel Materials Through 50 EFPY at the 1/4T Location(a) 1/4T R.G. 1.99, Predicted I (e) M Fluence(c) 1/4T RTNDT(U)(d) ART Material Rev. 2 CF (b)

RTNDT (x 1019 n/cm2, FF(c) (°F) (°F) (°F) (°F) (°F)

Position (°F)

E > 1.0 MeV)

Upper Shell B4711-1 1.1 73.5 0.0145 0.141 60 10.3 0.0 5.2 10.3 80.7 Upper Shell B4711-2 1.1 98.2 0.0145 0.141 60 13.8 0.0 6.9 13.8 87.6 Upper Shell B4711-3 1.1 82.6 0.0145 0.141 101 11.6 0.0 5.8 11.6 124.2 Intermediate Shell B4712-1 1.1 89.8 1.22 1.056 0 94.8 0.0 17.0 34.0 128.8 Intermediate Shell B4712-2 1.1 83.2 1.22 1.056 12 87.8 0.0 17.0 34.0 133.8 Using Credible Surveillance Data(f) 2.1 102.2 1.22 1.056 12 107.9 0.0 8.5 17.0 136.9(f)

Intermediate Shell B4712-3 1.1 73.7 1.22 1.056 10 77.8 0.0 17.0 34.0 121.8 Lower Shell B4713-1 1.1 83.0 1.20 1.050 8 87.2 0.0 17.0 34.0 129.2 Lower Shell B4713-2 1.1 82.4 1.20 1.050 8 86.6 0.0 17.0 34.0 128.6 Lower Shell B4713-3 1.1 82.6 1.20 1.050 10 86.8 0.0 17.0 34.0 130.8 Intermediate to Upper Shell 1.1 205.6 0.0167 0.154 -56 31.6 17.0 15.8 46.4 22.0 Circumferential Weld Seam 8-442 Intermediate to Lower Shell 1.1 91.4 1.18 1.046 -56 95.6 17.0 28.0 65.5 105.1 Circumferential Weld Seam 9-442 Upper Shell Longitudinal Weld Seams 1.1 208.7 0.0109 0.116 -56 24.3 17.0 12.1 41.8 10.0 1-442 A & C (60° / 300°)

Upper Shell Longitudinal Weld Seam 1.1 208.7 0.00602 0.0768 -56 16.0 17.0 8.0 37.6 -2.4 1-442 B (180°)

Intermediate Shell Longitudinal Weld 1.1 189.1 0.876 0.963 -56 182.1 17.0 28.0 65.5 191.6 Seams 2-442 B & C (120° / 240°)

Intermediate Shell Longitudinal Weld 1.1 189.1 0.442 0.773 -56 146.1 17.0 28.0 65.5 155.6 Seam 2-442 A (0°)

Lower Shell Longitudinal Weld Seams 1.1 208.6 0.864 0.959 -56 200.1 17.0 28.0 65.5 209.6 3-442 A & C (60° / 300°)

Using Credible Surveillance Data for 2.1 200.7 0.864 0.959 -56 192.5 17.0 14.0 44.0 180.5(f)

Weld Heat # 21935/12008(f)

Lower Shell Longitudinal Weld Seam 1.1 208.6 0.440 0.772 -56 161.0 17.0 28.0 65.5 170.5 3-442 B (180°)

Using Credible Surveillance Data for 2.1 200.7 0.440 0.772 -56 154.9 17.0 14.0 44.0 142.9(f)

Weld Heat # 21935/12008(f)

Notes for Table 6-2 contained on the next page.

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Westinghouse Non-Proprietary Class 3 6-4 Notes for Table 6-2:

(a) The Regulatory Guide 1.99, Revision 2 [Ref. 4] methodology was utilized in the calculation of the ART values.

(b) CFs are taken from Table 5-3.

(c) Fluence and FFs are taken from Table 6-1.

(d) RTNDT(U) and I values are taken from Table 3-1.

(e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and = 8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT for either base metals or welds, with or without surveillance data.

(f) The credibility evaluation in Appendix A of this report determined that the surveillance data for Intermediate Shell Plate B4712-2 are deemed credible.

The surveillance weld data for Heat # 21935/12008 from the Diablo Canyon Unit 2 surveillance program, applicable to the Lower Shell Longitudinal Weld Seams 3-442 A, B, & C are deemed credible in WCAP-17315-NP [Ref. 15]. Therefore, the Position 2.1 CF can be used with a reduced margin term in lieu of the Position 1.1 CF.

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Westinghouse Non-Proprietary Class 3 6-5 Table 6-3 Adjusted Reference Temperature Evaluation for the Salem Unit 2 Reactor Vessel Materials Through 50 EFPY at the 3/4T Location(a) 3/4T R.G. 1.99, Predicted I (e) M Fluence(c) 3/4T RTNDT(U)(d) ART Material Rev. 2 CF (b)

RTNDT (x 1019 n/cm2, FF(c) (°F) (°F) (°F) (°F) (°F)

Position (°F)

E > 1.0 MeV)

Upper Shell B4711-1 1.1 73.5 0.00517 0.069 60 5.0 0.0 2.5 5.0 70.1 Upper Shell B4711-2 1.1 98.2 0.00517 0.069 60 6.7 0.0 3.4 6.7 73.5 Upper Shell B4711-3 1.1 82.6 0.00517 0.069 101 5.7 0.0 2.8 5.7 112.3 Intermediate Shell B4712-1 1.1 89.8 0.434 0.768 0 69.0 0.0 17.0 34.0 103.0 Intermediate Shell B4712-2 1.1 83.2 0.434 0.768 12 63.9 0.0 17.0 34.0 109.9 Using Credible Surveillance Data(f) 2.1 102.2 0.434 0.768 12 78.5 0.0 8.5 17.0 107.5(f)

Intermediate Shell B4712-3 1.1 73.7 0.434 0.768 10 56.6 0.0 17.0 34.0 100.6 Lower Shell B4713-1 1.1 83.0 0.426 0.763 8 63.3 0.0 17.0 34.0 105.3 Lower Shell B4713-2 1.1 82.4 0.426 0.763 8 62.8 0.0 17.0 34.0 104.8 Lower Shell B4713-3 1.1 82.6 0.426 0.763 10 63.0 0.0 17.0 34.0 107.0 Intermediate to Upper Shell 1.1 205.6 0.00593 0.076 -56 15.6 17.0 7.8 37.4 -3.0 Circumferential Weld Seam 8-442 Intermediate to Lower Shell 1.1 91.4 0.419 0.759 -56 69.3 17.0 28.0 65.5 78.9 Circumferential Weld Seam 9-442 Upper Shell Longitudinal Weld Seams 1.1 208.7 0.00387 0.055 -56 11.5 17.0 5.8 35.9 -8.5 1-442 A & C (60° / 300°)

Upper Shell Longitudinal Weld Seam 1.1 208.7 0.00214 0.035 -56 7.2 17.0 3.6 34.8 -14.0 1-442 B (180°)

Intermediate Shell Longitudinal Weld 1.1 189.1 0.311 0.680 -56 128.5 17.0 28.0 65.5 138.1 Seams 2-442 B & C (120° / 240°)

Intermediate Shell Longitudinal Weld 1.1 189.1 0.157 0.513 -56 97.0 17.0 28.0 65.5 106.5 Seam 2-442 A (0°)

Lower Shell Longitudinal Weld Seams 1.1 208.6 0.307 0.676 -56 141.1 17.0 28.0 65.5 150.6 3-442 A & C (60° / 300°)

Using Credible Surveillance Data for 2.1 200.7 0.307 0.676 -56 135.7 17.0 14.0 44.0 123.8(f)

Weld Heat # 21935/12008(f)

Lower Shell Longitudinal Weld Seam 1.1 208.6 0.156 0.512 -56 106.8 17.0 28.0 65.5 116.3 3-442 B (180°)

Using Credible Surveillance Data for 2.1 200.7 0.156 0.512 -56 102.8 17.0 14.0 44.0 90.8(f)

Weld Heat # 21935/12008(f)

Notes for Table 6-3 contained on the next page.

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Westinghouse Non-Proprietary Class 3 6-6 Notes for Table 6-3:

(a) The Regulatory Guide 1.99, Revision 2 [Ref. 4] methodology was utilized in the calculation of the ART values.

(b) CFs are taken from Table 5-3.

(c) Fluence and FFs are taken from Table 6-1.

(d) RTNDT(U) and I values are taken from Table 3-1.

(e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and = 8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT for either base metals or welds, with or without surveillance data.

(f) The credibility evaluation in Appendix A of this report determined that the surveillance data for intermediate shell plate B4712-2 are deemed credible.

The surveillance weld data for Heat # 21935/12008 from the Diablo Canyon Unit 2 surveillance program, applicable to the Lower Shell Longitudinal Weld Seams 3-442 A, B, & C are deemed credible in WCAP-17315-NP [Ref. 15]. Therefore, the Position 2.1 CF can be used with a reduced margin term in lieu of the Position 1.1 CF.

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Westinghouse Non-Proprietary Class 3 6-7 Table 6-4 Limiting ART Values for Salem Unit 2 at 50 EFPY Limiting ART Values(a)

Limiting Material

(°F)

Intermediate Shell Longitudinal 1/4T location 191.6(b)

Weld Seams 2-442 B & C Intermediate Shell Longitudinal 3/4T location 138.1(b)

Weld Seams 2-442 B & C Notes:

(a) Values are the limiting values from Tables 6-2 and 6-3.

(b) Note, the ART value calculated for Lower Shell Longitudinal Weld Seams 3-442 A & C have higher ART values at both the 1/4T and 3/4T locations when a Position 1.1 CF is used. Because credible surveillance data exists from the Diablo Canyon Unit 2 surviellance program, the lower 1/4T and 3/4T ART values, calculated with a Position 2.1 CF, supersede those values.

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Westinghouse Non-Proprietary Class 3 7-1 7 P-T LIMIT CURVE APPLICABILITY EVALUATION The existing P-T limit curves, implemented in Technical Specifications Figures 3.4-2 and 3.4-3, were calculated in WCAP-15566 [Ref. 19] and are valid through 32 EFPY. In support of license renewal, new P-T limit curves were generated in WCAP-16982-NP [Ref. 1] valid through 50 EFPY. In order to support the implementation of the 50 EFPY P-T limit curves, the applicability of the P-T limit curves is evaluated by comparing the 50 EFPY limiting 1/4T and 3/4T ART values contained in Tables 6-2 and 6-3 with those used in WCAP-16982-NP.

This comparison is provided in Table 7-1.

Table 7-1 Summary of the Limiting ART Values 1/4T Limiting ART 3/4T Limiting ART

(°F) (°F)

WCAP-16982-NP Table 6-4 WCAP-16982-NP Table 6-4 209 191.6 150 138.1 Table 7-1 shows that the limiting ART values at the 1/4T and 3/4T locations will not exceed the ART values used in the P-T limit curves of WCAP-16982-NP. Furthermore, the limiting initial RTNDT value of 15°F for the closure head flange and the vessel flange remains appropriate, and the P-T limit curves flange notch and bolt-up temperature require no change or further consideration. Therefore, the EOLE (50 EFPY) P-T limit curves generated in WCAP-16982-NP can be implemented as planned.

The enable temperature presented in WCAP-16982-NP is based on the same ART values as the P-T limit curves. Since the P-T limit curves remain valid, the enable temperature also remains valid. Thus, from WCAP-16982-NP, the minimum required enable temperature without margins for instrument uncertainty is a coolant temperature equal to 274°F through EOLE. With margins for instrument uncertainty (+18°F),

the minimum required enable temperature is a coolant temperature equal to 292°F.

Inlet and Outlet Nozzles P-T Limit Curves NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. 9] requires that the P-T limit curves account for the higher stresses in the nozzle corner region due to the potential for more restrictive P-T limits, even if the RTNDT for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.

PWROG-15109-NP-A [Ref. 20] addresses this concern generically for the U.S. pressurized water reactor (PWR) operating fleet. The results of PWROG-15109-NP-A demonstrate that P-T limit curves developed with current NRC-approved methods (e.g., WCAP-14040-A [Ref. 21]) bound the generic nozzle P-T limit curves. This document has been approved by the NRC in Refence 22 as an acceptable means to address the concerns of RIS 2014-11. The results and conclusions of PWROG-15109-NP-A are applicable as long as the plant-specific Salem Unit 2 fluence of the nozzle corners remains less than the screening criterion of WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 7-2 4.28 x 1017 n/cm2 (E > 1.0 MeV), as described in PWROG-15109-NP-A. Section 2 of this report demonstrates Salem Unit 2 adherence to this screening criterion, thus PWROG-15109-NP-A is applicable.

In conclusion, PWROG-15109-NP-A demonstrates that the nozzles will not be limiting with respect to the P-T limit curves at Salem Unit 2. Therefore, the concerns of RIS 2014-11 are adequately addressed.

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Westinghouse Non-Proprietary Class 3 8-1 8 REFERENCES

1. Westinghouse Report WCAP-16982-NP, Revision 3, Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity, December 2020.

2. Regulatory Guide 1.190, Revision 0, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, March 2001. [Agencywide Documents Access and Management System [ADAMS] Accession Number ML010890301]

3. Westinghouse Report WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018.

4. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.

Nuclear Regulatory Commission, May 1988. [ ADAMS Accession Number ML003740284]

5. Radiation Safety Information Computational Center (RSICC) Data Library Collection DLC-185, BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, Oak Ridge National Laboratory (ORNL), July 1999.

6. ORNL Report ORNL/TM-13205, Pool Critical Assembly Pressure Vessel Facility Benchmark, (NUREG/CR-6454), July 1997.

7. ORNL Report ORNL/TM-13204, H.B. Robinson-2 Pressure Vessel Benchmark, (NUREG/CR-6453), October 1997 (published February 1998).

8. Code of Federal Regulations 10 CFR 50, Appendix G, Fracture Toughness Requirements, U.S.

Nuclear Regulatory Commission, Federal Register, November 29, 2019.

9. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S. Nuclear Regulatory Commission, October 2014. [ADAMS Accession No. ML14149A165]

10. Combustion Engineering Owners Group Report CE NPSD-1119, Revision 01, Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content, July 1998.

11. Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

12. Westinghouse Report WCAP-15423, Revision 0, Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, September 2000.

13. Westinghouse Report WCAP-18502-NP, Revision 1, Salem Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, July 2020.

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Westinghouse Non-Proprietary Class 3 8-2

14. Electric Power Research Institute (EPRI) 2011 Technical Report, BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. Palo Alto, CA: July 2011. 1022835.

15. Westinghouse Report WCAP-17315-NP, Revision 0, Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations, July 2011.

16. Westinghouse Report WCAP-15692, Revision 0, Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, August 2001.

17. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC Presentation, Generic Letter 92-01 and RPV Integrity Assessment, Status, Schedule, and Issues, NRC/Industry Workshop on RPV Integrity Issues, February 1998. [ADAMS Accession Number ML110070570]

18. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subarticle NB-2300, Fracture Toughness Requirements for Material.

19. Westinghouse Report WCAP-15566, Revision 1, Salem Unit 2 Heatup and Cooldown Curves for Normal Operation, February 2001.

20. PWR Owners Group Report PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020. [ADAMS Accession Number ML20024E573]

21. Westinghouse Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

22. NRC Safety Evaluation Final Safety Evaluation by the Office of Nuclear Reactor Regulation for Pressurized Water Reactor Owners Group Topical Report PWROG-15109-NP, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation EPID L-2018-TOP-0009, October 31, 2019. [ADAMS Accession Numbers ML19301D063 & ML19301D160]

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A CREDIBILITY RE-EVALUATION OF THE SALEM UNIT 2 SURVEILLANCE PROGRAM Regulatory Guide 1.99, Revision 2 [Ref. A-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describe the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date, there have been four surveillance capsules removed and tested from the Salem Unit 2 reactor vessel.

In addition, surveillance data from a sister plant is utilized herein to perform neutron radiation embrittlement calculations.

Table A-1 reviews the five criteria of Regulatory Guide 1.99, Revision 2. The following subsections evaluate each of these five criteria for Salem Unit 2 in order to determine the credibility of the surveillance data for use in neutron radiation embrittlement calculations.

Table A-1 Identification of Credibility Criteria Affected by Fluence Change Criterion Description No.

Materials in the capsules should be those judged most likely to be controlling with regard 1

to radiation embrittlement.

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated 2 conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two 3 or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. A-4].

The irradiation temperature of the Charpy specimens in the capsule should match the vessel 4

wall temperature at the cladding/base metal interface within +/- 25°F.

The surveillance data for the correlation monitor material in the capsule should fall within 5

the scatter band of the database for that material.

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Westinghouse Non-Proprietary Class 3 A-2 Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR 50, Fracture Toughness Requirements, [Ref. A-2] as follows:

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

The Salem Unit 2 reactor vessel consists of the following beltline region materials, which likely would have been considered at the time the surveillance program was designed and licensed:

Intermediate Shell Plates B4712-1, B4712-2, and B4712-3

Lower Shell Plates B4713-1, B4713-2, and B4713-3

Intermediate to Lower Shell Circumferential Weld Seam 9-442 (Heat # 90099, Linde 0091 flux, Lot # 3977)

Intermediate Shell Longitudinal Weld Seams 2-442 A, 2-442 B, and 2-442 C (Heat # 13253/20291, Linde 1092 flux, Lot # 3833)

Lower Shell Longitudinal Weld Seams 3-442 A, 3-442 B, and 3-442 C (Heat # 21935/12008, Linde 1092 flux, Lot # 3889)

Intermediate Shell Plate B4712-2 has the highest initial RTNDT of the beltline plates and was selected for the surveillance program. The surveillance weld was fabricated with weld wire Heat # 13253 and Linde 1092 flux, Lot # 3833/3774 utilizing the same fabrication practice as that used to fabricate the actual vessel beltline welds. This weld was made of one of the heats and with the same type flux as was used in the intermediate shell longitudinal weld seams.

Thus, the Salem Unit 2 surveillance program meets the intent of this criterion.

Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

The surveillance capsule analysis report, WCAP-15692 [Ref. A-5], was reviewed and it was determined that this criterion is met, consistent with the analysis in these documents.

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Westinghouse Non-Proprietary Class 3 A-3 Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. A-4].

The functional form of the least-squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these RTNDT values about this line is less than 28°F for welds and less than 17°F for the plates.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998 [Ref. A-3]. At this meeting, the NRC presented five cases. Of the five cases, Case 1 (Surveillance Data Available from Plant but No Other Source) most closely represents the situation for the Salem Unit 2 surveillance plate and weld materials. Case 5 (Surveillance Data from Other Sources Only) most closely represents the situation for the Salem Unit 2 Lower Shell Longitudinal Weld Seams 3-442 A, 3-442 B, and 3-442 C (Heat # 21935/12008) weld material.

Evaluation of the Salem Unit 2 Data Only (Case 1)

Following the NRC Case 1 guidelines, the Salem Unit 2 surveillance plates and weld metal (Heat # 13253) will be evaluated using the Salem Unit 2 data. Table A-2 provides the calculation of the interim CF for Salem Unit 2. Note that when evaluating the credibility of the surveillance weld data, the measured RTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2 Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. In addition, only Salem Unit 2 data are being considered; therefore, no temperature adjustment is required.

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Westinghouse Non-Proprietary Class 3 A-4 Table A-2 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Salem Unit 2 Surveillance Capsule Data Only Capsule Fluence(a) RTNDT(a) FF*RTNDT Material Capsule FF(b) FF2 (x 1019 n/cm2, (°F) (°F)

E > 1.0 MeV)

T 0.273 0.646 61.66 39.84 0.418 Intermediate Shell Plate U 0.581 0.848 66.54 56.43 0.719 B4712-2 X 1.13 1.034 93.82 97.02 1.069 (longitudinal)

Y 1.83 1.166 105.69 123.21 1.359 T 0.273 0.646 74.83 48.35 0.418 Intermediate Shell Plate U 0.581 0.848 98.26 83.33 0.719 B4712-2 X 1.13 1.034 125.15 129.42 1.069 (transverse)

Y 1.83 1.166 129.33 150.76 1.359 SUM: 728.36 7.130 CFB4712-2 = (FF

RTNDT) ÷ (FF2) = (728.36) ÷ (7.130) = 102.2°F T 0.273 0.646 153.17 98.97 0.418 U 0.581 0.848 185.94 157.68 0.719 Salem Unit 2 surveillance weld X 1.13 1.034 195.43 202.10 1.069 material Y 1.83 1.166 200.90 234.20 1.359 (Heat # 13253)

SUM: 692.95 3.565 CFSurv. Weld = (FF

RTNDT) ÷ (FF2) = (692.95) ÷ (3.565) = 194.4°F Notes:

(a) Fluence and measured RTNDT taken from Table 4-1.

(b) FF = fluence factor = f(0.28 - 0.10l*log(f)).

The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2 Position 2.1 is presented in Table A-3.

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Westinghouse Non-Proprietary Class 3 A-5 Table A-3 Surveillance Capsule Data Scatter about the Best-Fit Line Using Only Salem Unit 2 Surveillance Data

< 17°F Capsule CF(a) Measured Predicted Scatter (Base Fluence(b)

Material Capsule (Slopebest-fit) FF(c) RTNDT(b) RTNDT(d) RTNDT(e) Metal)

(x 1019 n/cm2,

(°F) (°F) (°F) (°F) < 28°F E > 1.0 MeV)

(Weld)

T 102.2 0.273 0.646 61.66 66.0 4.4 Yes Intermediate Shell Plate U 102.2 0.581 0.848 66.54 86.6 20.1 No B4712-2 X 102.2 1.13 1.034 93.82 105.6 11.8 Yes (longitudinal)

Y 102.2 1.83 1.166 105.69 119.1 13.4 Yes T 102.2 0.273 0.646 74.83 66.0 8.8 Yes Intermediate Shell Plate U 102.2 0.581 0.848 98.26 86.6 11.6 Yes B4712-2 X 102.2 1.13 1.034 125.15 105.6 19.5 No (transverse)

Y 102.2 1.83 1.166 129.33 119.1 10.2 Yes T 194.4 0.273 0.646 153.17 125.6 27.6 Yes Salem Unit 2 surveillance U 194.4 0.581 0.848 185.94 164.8 21.1 Yes weld material X 194.4 1.13 1.034 195.43 201.0 5.6 Yes (Heat # 13253)

Y 194.4 1.83 1.166 200.90 226.6 25.7 Yes Notes:

(a) CFs calculated in Table A-2.

(b) Fluence and measured RTNDT values are taken from Table 5-1.

(c) FF = fluence factor = f(0.28 - 0.10*log (f)).

(d) Predicted RTNDT = CF x FF.

(e) Scatter RTNDT = Absolute Value [Predicted RTNDT - Measured RTNDT].

From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table A-3 indicates that six of the eight (75%) surveillance data points fall inside the +/- 1 of 17°F scatter band for surveillance base metals. Therefore, all the plate data are deemed credible per the third criterion.

Table A-3 indicates that all surveillance data points fall inside the +/- 1 of 28°F scatter band for surveillance weld materials. 100% of the data are bounded; therefore, the surveillance weld data are deemed credible per the third criterion. Note that, while credible, the data for surveillance weld Heat # 13253 was determined not to be relevant to the welds in the reactor vessel; therefore, the data are not used for embrittlement calculations of the Salem Unit 2 reactor vessel.

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Westinghouse Non-Proprietary Class 3 A-6 Evaluation of Weld Data from All Sources (Case 5)

Next, data from all sources are considered in order to evaluate the credibility of Heat # 21935/12008 used in the Salem Unit 2 Lower Shell Longitudinal Weld Seams 3-442 A, 3-442 B, and 3-442 C using the NRC Case 5 guidelines.

WCAP-17315-NP [Ref. A-6] already concluded that the data for the Diablo Canyon Unit 2 surveillance weld are credible, therefore, the data are not re-analyzed separately herein.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.

The capsule specimens are located in the reactor between the core barrel and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pad. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F. Hence, this criterion is met.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The Salem Unit 2 surveillance program does not include correlation monitor material. Therefore, this criterion is not applicable to Salem Unit 2.

CONCLUSION:

Based on the preceding responses to the five criteria of Regulatory Gide 1.99, Revision 2 [Ref. A-1],

Section B, the Salem Unit 2 surveillance data are credible.

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Westinghouse Non-Proprietary Class 3 A-7 REFERENCES A-1 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.

Nuclear Regulatory Commission, May 1988. [ADAMS Accession Number ML003740284]

A-2 Code of Federal Regulations 10 CFR 50, Appendix G, Fracture Toughness Requirements, U.S.

Nuclear Regulatory Commission, Federal Register, November 29, 2019.

A-3 K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC Presentation, Generic Letter 92-01 and RPV Integrity Assessment, Status, Schedule, and Issues, NRC/Industry Workshop on RPV Integrity Issues, February 1998. [ADAMS Accession Number ML110070570]

A-4 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society for Testing and Materials, 1982.

A-5 Westinghouse Report WCAP-15692, Revision 0, Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, August 2001.

A-6 Westinghouse Report WCAP-17315-NP, Revision 0, Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations, July 2011.

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Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION B.1 PRESSURIZED THERMAL SHOCK A limiting condition on RPV integrity known as pressurized thermal shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the RPV under the following conditions: severe overcooling of the inside surface of the vessel wall followed by high pressurization, significant degradation of vessel material toughness caused by radiation embrittlement, and the presence of a critical-size defect anywhere within the vessel wall.

In 1985, the U.S. NRC issued a formal ruling on PTS (10 CFR 50.61 [Ref. B-1]) that established screening criteria on pressurized water reactor (PWR) vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTS.

RTPTS screening values were set by the NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seals for plant operation to the end-of-plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end of license.

The NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RTPTS) values consistent with the methods given in Regulatory Guide 1.99, Revision 2

[Ref. B-2].

These accepted methods were used with the surface fluence values of Section 2 to calculate the following RTPTS values for the Salem Unit 2 RPV materials. The RTPTS calculations are presented in Table B-1 for Salem Unit 2.

PTS Conclusion All of the reactor vessel materials for Salem Unit 2 are below the RTPTS screening criteria values of 270°F for plates, forgings, and longitudinal welds, and 300°F for circumferentially oriented welds (per 10 CFR 50.61) at 50 EFPY.

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Westinghouse Non-Proprietary Class 3 B-2 Table B-1 RTPTS Calculations for Salem Unit 2 Reactor Vessel Materials at 50 EFPY(a)

Surface Fluence(c) Predicted Surf. RTNDT(U)(e) U (f) M RTPTS Material CF(b) (x 1019 n/cm2, RTNDT FF(d) (°F) (°F) (°F) (°F) (°F)

E > 1.0 MeV) (°F)

Upper Shell B4711-1 73.5 0.0244 0.194 60 14.3 0.0 7.1 14.3 88.6 Upper Shell B4711-2 98.2 0.0244 0.194 60 19.1 0.0 9.5 19.1 98.2 Upper Shell B4711-3 82.6 0.0244 0.194 101 16.0 0.0 8.0 16.0 133.1 Intermediate Shell B4712-1 89.8 2.05 1.196 0 107.4 0.0 17.0 34.0 141.4 Intermediate Shell B4712-2 83.2 2.05 1.196 12 99.5 0.0 17.0 34.0 145.5 Using Credible Surveillance Data(g) 102.2 2.05 1.196 12 122.2 0.0 8.5 17.0 151.2(g)

Intermediate Shell B4712-3 73.7 2.05 1.196 10 88.1 0.0 17.0 34.0 132.1 Lower Shell B4713-1 83.0 2.01 1.190 8 98.8 0.0 17.0 34.0 140.8 Lower Shell B4713-2 82.4 2.01 1.190 8 98.1 0.0 17.0 34.0 140.1 Lower Shell B4713-3 82.6 2.01 1.190 10 98.3 0.0 17.0 34.0 142.3 Intermediate to Upper Shell Circumferential 205.6 0.0280 0.211 -56 43.4 17.0 21.7 55.1 42.5 Weld Seam 8-442 Intermediate to Lower Shell Circumferential 91.4 1.98 1.186 -56 108.4 17.0 28.0 65.5 118.0 Weld Seam 9-442 Upper Shell Longitudinal Weld Seams 208.7 0.0183 0.163 -56 34.0 17.0 17.0 48.1 26.0 1-442 A & C (60°/ 300°)

Upper Shell Longitudinal Weld Seam 208.7 0.0101 0.110 -56 23.0 17.0 11.5 41.1 8.1 1-442 B (180°)

Intermediate Shell Longitudinal Weld Seams 189.1 1.47 1.107 -56 209.3 17.0 28.0 65.5 218.8 2-442 B & C (120°/ 240°)

Intermediate Shell Longitudinal Weld Seam 189.1 0.741 0.916 -56 173.2 17.0 28.0 65.5 182.7 2-442 A (0°)

Lower Shell Longitudinal Weld Seams 208.6 1.45 1.103 -56 230.1 17.0 28.0 65.5 239.6 3-442 A & C (60°/ 300°)

Using Credible Surveillance Data 200.7 1.45 1.103 -56 221.4 17.0 14.0 44.0 209.4(g) for Weld Heat # 21935/12008(g)

Lower Shell Longitudinal Weld Seam 208.6 0.738 0.915 -56 190.8 17.0 28.0 65.5 200.3 3-442 B (180°)

Using Credible Surveillance Data 200.7 0.738 0.915 -56 183.6 17.0 14.0 44.0 171.6(g) for Weld Heat # 21935/12008(g)

Notes for Table B-1 contained on the next page.

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Westinghouse Non-Proprietary Class 3 B-3 Notes for Table B-1:

(a) The 10 CFR 50.61 [Ref. B-1] methology was utilized in the calculation of the RTPTS values.

(b) CFs are taken from Table 5-3.

(c) Fluence values are taken from Table 2-2.

(d) FF = fluence factor = f(0.28 - 0.10*log (f)).

(e) RTNDT(U) and I values are taken from Table 3-1.

(f) Per 10 CFR 50.61, the base metal = 17°F when surveillance data are non-credible or not used to determine the CF, and = 8.5°F when credible surveillance data are used to determine the CF. Also, per 10 CFR 50.61, the weld metal = 28°F when surveillance data are non-credible or not used to determine the CF, and = 14°F when credible surveillance data are used to determine the CF. However, need not exceed 0.5*RTNDT.

(g) The credibility evaluation in Appendix A of this report determined that the surveillance data for Intermediate Shell Plate B4712-2 are deemed credible.

The surveillance weld data for Heat # 21935/12008 from the Diablo Canyon Unit 2 surveillance program, applicable to the Lower Shell Longitudinal Weld Seams 3-442 A, B, & C, are deemed credible in WCAP-17315-NP [Ref. B-3]. Therefore, the Position 2.1 CF can be used with a reduced margin term in lieu of the Position 1.1 CF.

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Westinghouse Non-Proprietary Class 3 B-4 B.2 EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION The emergency response guideline (ERG) limits, HF04BG [Ref. B-4], were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event. Generic categories of limits were developed for the guidelines based on the limiting inside surface RTNDT. These generic categories were conservatively generated for the Westinghouse Owners Group (WOG - now known as Pressurized Water Reactor Owners Group [PWROG]) to be applicable to all Westinghouse plants.

The highest value of RTNDT, for which the generic category of ERG limits was developed, is 250°F for a longitudinal flaw and 300°F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250°F for a longitudinal flaw or 300°F for a circumferential flaw, plant-specific ERG P-T limits must be developed.

The ERG category is determined by the magnitude of the limiting RTNDT value, which is calculated the same way as the RTPTS values are calculated in Section B.1 of this report. The material with the highest RTNDT defines the limiting material. Table B-2 identifies ERG category limits and the limiting material RTNDT values at EOLE for Salem Unit 2.

Table B-2 Evaluation of Salem Unit 2 ERG Limit Category ERG Pressure-Temperature Limits [Ref. B-4]

Applicable RTNDT Value(a) ERG P-T Limit Category RTNDT < 200°F Category I 200°F < RTNDT < 250°F Category II 250°F < RTNDT < 300°F Category III b Limiting RTNDT Value(b)

Limiting Reactor Vessel Material RTNDT Value @ EOLE Salem Unit 2 Intermediate Shell Longitudinal Weld 200°F < RTNDT < 250°F Seams 2-442 B & C Notes:

(a) Longitudially oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.

(b) Values taken from Table B-1.

Per the ERG limit guidance document [Ref. B-4], some vessels do not change categories for operation through the end of license. However, when a vessel does change ERG categories between the beginning and end of operation, a plant-specific assessment must be performed to determine at what operating time the category changes. Thus, the ERG classification need not be changed until the operating cycle during which the maximum vessel value of actual or estimated real-time RTNDT exceeds the limit on its current ERG category.

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Westinghouse Non-Proprietary Class 3 B-5 Conclusion of ERG P-T Limit Categorization Per Table B-2, the limiting material for Salem Unit 2 assumes a longitudinally oriented flaw and has an RTNDT value above 200°F but below 250°F at EOLE. Therefore, Salem Unit 2 is limited to ERG Category II through EOLE and will not need plant-specific ERG P-T limits.

The Salem Unit 2 RTPTS will exceed 200°F, thus the plants ERG category changes from Category I to II, when the fluence of the Intermediate Shell Longitudinal Weld Seams 2-442 B & C exceeds 1.03 x 1019 n/cm2 (E > 1.0 MeV). Interpolations of the fluences in Table 2-2 show that this will occur at 34.3 EFPY. Therefore, Salem Unit 2 must switch from ERG Category I to Category II prior to the cycle in which 34.3 EFPY is reached. Salem Unit 2 will then remain ERG Category II through EOLE (50 EFPY).

REFERENCES B-1 Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

B-2 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.

Nuclear Regulatory Commission, May 1988. [ADAMS Accession Number ML003740284]

B-3 Westinghouse Report WCAP-17315-NP, Revision 0, Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations, July 2011.

B-4 Westinghouse Owners Group Report HF04BG, Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 3, March 2014.

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Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS C.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn from service to-date are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [Ref. C-1]. One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least-squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in this report.

C.2 SENSOR REACTION RATE DETERMINATIONS In this section, the results of the evaluations of the in-vessel neutron sensor sets withdrawn and analyzed to-date as part of the reactor vessel materials surveillance program are presented.

Eight irradiation capsules attached to the thermal shield were included in the reactor design to constitute the reactor vessel surveillance program. The capsules were located at azimuthal angles of 4° (Capsule S),

176° (Capsule V), 184° (Capsule W), and 356° (Capsule Z) that are 4° from the core cardinal axes and 40° (Capsule T), 140° (Capsule U), 220° (Capsule X), and 320° (Capsule Y) that are 40° from the core cardinal axes. The irradiation history of each of these eight in-vessel surveillance capsules is summarized as follows:

Capsule Location Irradiation History T 40° Cycle 1 (withdrawn for analysis)

U 40° Cycles 1-3 (withdrawn for analysis)

X 40° Cycles 1-6 (withdrawn for analysis)

Y 40° Cycles 1-11 (withdrawn for analysis)

S 4° In the reactor V 4° In the reactor W 4° In the reactor Z 4° In the reactor The azimuthal locations included in the above tabulation represent the first-octant-equivalent (FOE) azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of the surveillance capsules are summarized as follows:

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Westinghouse Non-Proprietary Class 3 C-2 Sensor Material Reaction of Capsule T Capsule U Capsule X Capsule Y Interest Copper 63 Cu(n,) 60Co X X X X 54 Iron Fe(n,p) 54Mn X X X X 58 Nickel Ni(n,p) 58Co X X X X 238 Uranium-238 U(n,f) 137Cs X X X X 237 Neptunium-237 Np(n,f) 137Cs X X X X Cobalt-Aluminum(a) 59 Co(n,) 60Co X X X(b) X Notes:

(a) The cobalt-aluminum measurements include both bare wire and cadmium-covered sensors.

(b) The bare cobalt-aluminum measurement for Capsule X was determined to be statistically different than similar measurement data obtained from the 4-loop, thermal-shield reactor plant database for 40° surveillance capsules. As a result, the Capsule X bare cobalt-aluminum measurement was not utilized in the least-squares adjustment for this capsule.

Pertinent physical and nuclear characteristics of the in-vessel surveillance capsule passive neutron sensors are listed in Table C-1.

The use of passive monitors, such as those listed above, does not yield a direct measure of the energy-dependent neutron fluence rate at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron fluence rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron fluence rate incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

The measured specific activity of each monitor

The physical characteristics of each monitor

The operating history of the reactor

The energy response of each monitor

The neutron energy spectrum at the monitor location Results from the radiometric counting of the neutron sensors from the in-vessel capsules are documented in References C-2-C-5, and re-evaluated in this appendix using the RAPTOR-M3G model described in previous sections. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by the in-vessel capsules was based on monthly power generation data from initial reactor criticality through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 C-3 representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The startup and shutdown dates for each cycle of operation used in the evaluations are given in Table C-2.

Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation:

=

0 1 ,

where R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref (rps/nucleus)

A = Measured specific activity (dps/g)

N0 = Number of target element atoms per gram of sensor F = Atom fraction of the target isotope in the target element Y = Number of product atoms produced per reaction Pj = Average core power level during irradiation Period j (MW)

Pref = Maximum or reference power level of the reactor (MW)

Cj = Calculated ratio of (E > 1.0 MeV) during irradiation Period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period

= Decay constant of the product isotope (1/sec) tj = Length of irradiation Period j (sec) td,j = Decay time following irradiation Period j (sec) and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variations of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology described in previous sections, accounts for the change in sensor reaction rates caused by variations in fluence rate induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 C-4 is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low-leakage fuel management, the additional Cj term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low-leakage to low-leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another. The fuel-cycle-specific neutron fluence rate values are used to compute cycle-dependent Cj values at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the fission sensor reaction rates are summarized as follows:

Correction Capsule T Capsule U Capsule X Capsule Y 235 U impurity/Pu build-in 0.8735 0.8616 0.8405 0.8148 238 U(,f) 0.9555 0.9558 0.9562 0.9564 Net 238U correction 0.8346 0.8235 0.8037 0.7793 237 Np(,f) 0.9840 0.9840 0.9841 0.9841 These factors were applied in a multiplicative fashion to the decay-corrected uranium and neptunium fission sensor reaction rates.

Additionally, radial gradient correction factors are applied to all samples which are not located at the center of the surveillance capsule. The radial correction factors are computed by calculating the reaction specific responses at the capsule center location and dividing by the corresponding response at the specified radial location. The following table summarizes the radial gradient correction factors.

Dosimeter Radius Gradient Correction Factor(a)

Reaction Location (cm) Capsule T Capsule U Capsule X Capsule Y 63 Cu(n,) 60Co Front 211.18 0.960 0.960 0.960 0.959 54 54 Fe(n,p) Mn Middle 211.68 1.054 1.054 1.054 1.055 58 Ni(n,p) 58Co Rear 212.18 1.170 1.171 1.172 1.173 238 137 U(n,f) Cs Center 211.41 -- -- -- --

237 Np(n,f) 137Cs Center 211.41 -- -- -- --

59 60 Co(n,) Co Rear 212.18 0.978 0.977 0.977 0.977 59 Co(Cd)(n,) 60Co Rear 212.18 1.161 1.160 1.160 1.160 Note:

(a) Gradient correction factors for the 238U and 237Np sensors are not required since they are located at the center of the capsule.

Results of the sensor reaction rate determinations for the in-vessel capsules are given in Table C-3 through Table C-6.

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Westinghouse Non-Proprietary Class 3 C-5 C.3 LEAST-SQUARES EVALUATION OF SENSOR SETS Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations, resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best estimates for key exposure parameters such as (E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties. For example:

+/- = +/- +/-

relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross section, ig, each with an uncertainty . The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least-squares evaluation of the surveillance capsule dosimetry, the FERRET Code [Ref. C-6] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters ( (E > 1.0 MeV) and dpa) along with associated uncertainties for the in-vessel capsules analyzed to-date.

The application of the least-squares methodology requires the following input:

1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2. The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the plant-specific application of the least-squares methodology, the calculated neutron spectrum was obtained from the results of the neutron transport calculations described in previous sections of this report.

The sensor reaction rates were derived from the measured specific activities using the procedures described in Section C.2. The dosimetry reaction cross sections and uncertainties were obtained from the Sandia National Laboratories Radiation Metrology Laboratory (SNLRML) dosimetry cross-section library

[Ref. C-7]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E1018-09, Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (IIB) [Ref. C-8].

The uncertainties associated with the measured reaction rates, dosimetry cross sections, and calculated neutron spectrum were input to the least-squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIIA) [Ref. C-9].

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Westinghouse Non-Proprietary Class 3 C-6 The following provides a summary of the uncertainties associated with the least-squares evaluation of the surveillance capsule sensor sets withdrawn and analyzed to-date.

C.4 REACTION RATE UNCERTAINTIES The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least-squares evaluation:

Reaction Uncertainty 63 Cu(n,) 60Co 5%

54 Fe(n,p) 54Mn 5%

58 Ni(n,p) 58Co 5%

238 U(n,f) 137Cs 10%

237 Np(n,f) 137Cs 10%

59 Co(n,) 60Co 5%

These uncertainties are given at the 1 level.

C.5 DOSIMETRY CROSS-SECTION UNCERTAINTIES The reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library.

This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multigroup structure for use in least-squares adjustment applications. These cross sections were compiled from the most recent cross-section evaluations, and they have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.

For sensors included in the plant-specific reactor vessel surveillance program, the following uncertainties in the fission spectrum averaged cross sections are provided in the SNLRML documentation package.

Reaction Uncertainty 63 Cu(n, ) 60Co 4.08-4.16%

54 Fe(n,p) 54Mn 3.05-3.11%

58 Ni(n,p) 58Co 4.49-4.56%

238 U(n,f) 137Cs (Cd) 0.54-0.64%

237 Np(n,f) 137Cs (Cd) 10.32-10.97%

59 Co(n,) 60Co 0.79-3.59%

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Westinghouse Non-Proprietary Class 3 C-7 These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

C.6 CALCULATED NEUTRON SPECTRUM The neutron spectra input to the least-squares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape).

Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

Using the uncertainties associated with the reaction rates obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

= 2 +

where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg specify additional random group-wise uncertainties that are correlated with a correlation matrix given by:

= [1 ] +

where

( )2

=

2 2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range ( specifies the strength of the latter term).

The value of is 1.0 when g = g and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the calculated spectra was as follows:

Flux Normalization Uncertainty (Rn) 15%

Flux Group Uncertainties (Rg, Rg)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlation ()

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 C-8 Flux Group Correlation Range ()

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 C.7 COMPARISONS OF MEASUREMENTS AND CALCULATIONS This section provides comparisons of the measurement results from each of the sensor set irradiations with corresponding analytical predictions at the measurement locations. These comparisons are provided on two levels. In the first level, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. This level of comparison is not impacted by the least-squares evaluations of the sensor sets. In the second level, calculated values of neutron exposure rates in terms of fast neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are compared with the best-estimate exposure rates obtained from the least-squares evaluation.

In Table C-7, comparisons of M/C ratios are listed for the threshold sensors contained in the in-vessel capsules. From Table C-7, it is noted that for the individual threshold sensors, the average M/C ratio ranges from 0.95 to 1.07 with an overall average of 1.01 and an associated standard deviation of 7.0%. In this case, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Table C-8, best-estimate-to-calculation (BE/C) ratios for fast neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate resulting from the least-squares evaluation of each dosimetry set. For the in-vessel capsules, the average BE/C ratio is 0.97 with an associated uncertainty of 4.7% for fast neutron fluence rate (E > 1.0 MeV) and 0.99 with an associated uncertainty of 4.3% for the iron atom displacement rate.

The M/C comparisons based on individual sensor reactions without recourse to the least-squares adjustment procedure are summarized as follows:

In-Vessel Capsules Reaction Avg. M/C % Unc. (1) 63 Cu(n,) 60Co 1.07 2.8%

54 54 Fe(n,p) Mn 0.95 5.8%

58 Ni(n,p) 58Co 0.96 5.0%

238 137 U(n,f) Cs 0.99 6.7%

237 Np(n,f) 137Cs 1.07 4.8%

Linear Average 1.01 7.0%

A similar comparison for exposure rate expressed in terms of neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate (dpa/s) are summarized as follows:

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Westinghouse Non-Proprietary Class 3 C-9 In-Vessel Capsules Parameter Avg. BE/C % Unc. (1)

Fast Neutron Fluence Rate (E > 1.0 MeV) 0.97 4.7%

Iron Atom Displacement Rate (dpa/s) 0.99 4.3%

These data comparisons show similar and consistent results, with the linear average M/C ratio of 1.01 in good agreement with the resultant least-squares BE/C ratios of 0.97 for neutron fluence rate (E > 1.0 MeV) and 0.99 for iron atom displacement rate. The comparisons demonstrate that the calculated results provided in previous sections of this report are validated within the context of the assigned 13% uncertainty and, further, show that the +/-20% (1) agreement between calculation and measurement required by [Ref. C-1]

is met.

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Westinghouse Non-Proprietary Class 3 C-10 Table C-1 Nuclear Parameters Used in the Evaluation of the In-Vessel Surveillance Capsule Neutron Sensors Reaction Atomic Target Product Fission of Weight(a) Atom Half-life(b)(c)(d) Yield(d)

Interest (g/g-atom) Fraction(b)(c) (days) (%)

63 Cu(n,) 60Co 63.546 0.6917 1925.28 N/A 54 Fe(n,p) 54Mn 55.845 0.05845 312.13 N/A 58 Ni(n,p) 58Co 58.6934 0.68077 70.86 N/A 238 U(n,f) 137Cs 238.051 1.00 10975.76 6.02 237 Np(n,f) 137Cs 237.048 1.00 10975.76 6.27 59 Co(n,) 60Co 58.933 0.0015 1925.28 N/A Notes:

(a) Atomic weight data were taken from the Chart of the Nuclides, 17th Edition, dated 2010 [Ref. C-10].

(b) Half-life and target atom fraction data for 63Cu(n,), 54Fe(n,p), and 58Ni(n,p), reactions were taken from ASTM Standard E1005-16 [Ref. C-11].

(c) The half-life for the 59Co(n,) reaction was taken from ASTM Standard E1005-16 [Ref. C-11]. The target atom fractions for the 59Co(n,), 238U(n,f), and 237Np(n,f) reactions are reflective of standard Westinghouse surveillance capsule dosimeter values.

(d) Half-life and fission yield data for the 238U(n,f) and 237Np(n,f) reactions were taken from ASTM Standard E1005-16

[Ref. C-11].

Table C-2 Startup and Shutdown Dates Cycle Startup Date Shutdown Date 1 06/03/1981 01/22/1983 2 07/31/1983 10/04/1984 3 04/13/1985 10/03/1986 4 12/23/1986 08/31/1988 5 11/26/1988 03/31/1990 6 06/24/1990 11/09/1991 7 04/30/1992 03/17/1993 8 06/20/1993 10/13/1994 9 02/04/1995 11/03/1995 10 08/31/1997 04/03/1999 11 05/28/1999 10/05/2000 WCAP-18571-NP July 2022 Revision 2

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Westinghouse Non-Proprietary Class 3 C-11 Table C-3 Measured Sensor Activities and Reaction Rates for Surveillance Capsule T Radially Corrected Average Measured Corrected Reaction Average Monitor Reaction Reaction Activity(a) Saturated Rate Reaction Identification Rate (Bq/g) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu - Top Mid Wire 63Cu(n,) 60Co 4.03E+04 3.10E+05 4.73E-17 Cu - Mid Wire 63Cu(n,) 60Co 4.06E+04 3.12E+05 4.76E-17 4.79E-17 4.79E-17 Cu - Bot Mid Wire 63Cu(n,) 60Co 4.16E+04 3.20E+05 4.88E-17 Fe - Top Wire 54Fe(n,p) 54Mn 7.67E+05 3.22E+06 5.11E-15 Fe - Top Mid Wire 54Fe(n,p) 54Mn 7.67E+05 3.22E+06 5.11E-15 Fe - Mid Wire 54Fe(n,p) 54Mn 7.35E+05 3.09E+06 4.90E-15 5.04E-15 5.04E-15 Fe - Bot Mid Wire 54Fe(n,p) 54Mn 7.56E+05 3.17E+06 5.04E-15 Fe - Bot Wire 54Fe(n,p) 54Mn 7.56E+05 3.17E+06 5.04E-15 Ni - Top Mid Wire 58Ni(n,p) 58Co 9.30E+05 4.89E+07 7.00E-15 Ni - Mid Wire 58Ni(n,p) 58Co 9.23E+05 4.85E+07 6.95E-15 6.99E-15 6.99E-15 Ni - Bot Mid Wire 58Ni(n,p) 58Co 9.34E+05 4.91E+07 7.03E-15 U-238 - Mid 238U(Cd)(n,f) 137Cs 1.23E+05 4.66E+06 3.06E-14 3.06E-14 2.55E-14 Np-237 - Mid 237Np(Cd)(n,f) 137Cs 9.49E+05 3.60E+07 2.26E-13 2.26E-13 2.22E-13 Co-Al - Top 59Co(n,) 60Co 7.05E+06 5.52E+07 3.60E-12 3.56E-12 3.56E-12 Co-Al - Bottom 59Co(n,) 60Co 6.90E+06 5.40E+07 3.53E-12 Co-Al - Top (Cd) 59Co(Cd)(n,) 60Co 2.84E+06 2.64E+07 1.72E-12 1.68E-12 1.68E-12 Co-Al - Bottom (Cd) 59Co(Cd)(n,) 60Co 2.69E+06 2.50E+07 1.63E-12 Note:

(a) Measured activities are decay corrected to January 16, 1984.

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Westinghouse Non-Proprietary Class 3 C-12 Table C-4 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U Radially Corrected Average Measured Corrected Reaction Average Monitor Reaction Reaction Activity(a) Saturated Rate Reaction Identification Rate (Bq/g) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu - Top Mid Wire 63Cu(n,) 60Co 7.17E+04 2.85E+05 4.35E-17 Cu - Mid Wire 63Cu(n,) 60Co 7.40E+04 2.95E+05 4.49E-17 4.48E-17 4.48E-17 Cu - Bot Mid Wire 63Cu(n,) 60Co 7.59E+04 3.02E+05 4.61E-17 Fe - Top Wire 54Fe(n,p) 54Mn 1.00E+06 2.67E+06 4.23E-15 Fe - Top Mid Wire 54Fe(n,p) 54Mn 1.01E+06 2.69E+06 4.27E-15 4.37E-15 4.37E-15 Fe - Bot Mid Wire 54Fe(n,p) 54Mn 1.06E+06 2.83E+06 4.48E-15 Fe - Bot Wire 54Fe(n,p) 54Mn 1.06E+06 2.83E+06 4.48E-15 Ni - Top Mid Wire 58Ni(n,p) 58Co 5.65E+06 4.05E+07 5.80E-15 Ni - Mid Wire 58Ni(n,p) 58Co 5.76E+06 4.13E+07 5.92E-15 5.93E-15 5.93E-15 Ni - Bot Mid Wire 58Ni(n,p) 58Co 5.90E+06 4.23E+07 6.06E-15 U-238 - Mid 238U(Cd)(n,f) 137Cs 2.65E+05 4.57E+06 3.00E-14 3.00E-14 2.47E-14 Np-237 - Mid 237Np(Cd)(n,f) 137Cs 1.94E+06 3.34E+07 2.10E-13 2.10E-13 2.07E-13 Co-Al - Top 59Co(n,) 60Co 1.17E+07 4.74E+07 3.09E-12 3.03E-12 3.03E-12 Co-Al - Bottom 59Co(n,) 60Co 1.12E+07 4.54E+07 2.96E-12 Co-Al - Top (Cd) 59Co(Cd)(n,) 60Co 4.57E+06 2.20E+07 1.43E-12 1.42E-12 1.42E-12 Co-Al - Bottom (Cd) 59Co(Cd)(n,) 60Co 4.45E+06 2.14E+07 1.40E-12 Note:

(a) Measured activities are decay corrected to February 2, 1987.

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Westinghouse Non-Proprietary Class 3 C-13 Table C-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X Radially Corrected Average Measured Corrected Reaction Average Monitor Reaction Reaction Activity(a) Saturated Rate Reaction Identification Rate (Bq/g) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu - Top Mid Wire 63Cu(n,) 60Co 1.11E+05 2.49E+05 3.80E-17 Cu - Mid Wire 63Cu(n,) 60Co 1.13E+05 2.54E+05 3.87E-17 3.87E-17 3.87E-17 Cu - Bot Mid Wire 63Cu(n,) 60Co 1.15E+05 2.58E+05 3.94E-17 Fe - Top Wire 54Fe(n,p) 54Mn 1.24E+06 2.52E+06 4.00E-15 Fe - Top Mid Wire 54Fe(n,p) 54Mn 1.15E+06 2.34E+06 3.71E-15 Fe - Mid Wire 54Fe(n,p) 54Mn 1.14E+06 2.32E+06 3.68E-15 3.77E-15 3.77E-15 Fe - Bot Mid Wire 54Fe(n,p) 54Mn 1.18E+06 2.40E+06 3.81E-15 Fe - Bot Wire 54Fe(n,p) 54Mn 1.13E+06 2.30E+06 3.65E-15 Ni - Top Mid Wire 58Ni(n,p) 58Co 7.90E+06 3.65E+07 5.23E-15 Ni - Mid Wire 58Ni(n,p) 58Co 7.97E+06 3.69E+07 5.28E-15 5.21E-15 5.21E-15 Ni - Bot Mid Wire 58Ni(n,p) 58Co 7.75E+06 3.58E+07 5.13E-15 U-238 - Mid 238U(Cd)(n,f) 137Cs 4.62E+05 3.67E+06 2.41E-14 2.41E-14 1.94E-14 Np-237 - Mid 237Np(Cd)(n,f) 137Cs 3.38E+06 2.68E+07 1.68E-13 1.68E-13 1.66E-13 Co-Al - Top 59Co(n,) 60Co 7.13E+06 1.63E+07 1.06E-12 9.80E-13 9.80E-13 Co-Al - Bottom 59Co(n,) 60Co 6.03E+06 1.38E+07 8.98E-13 Co-Al - Bottom (Cd) 59Co(Cd)(n,) 60Co 6.71E+06 1.82E+07 1.19E-12 1.19E-12 1.19E-12 Note:

(a) Measured activities are decay corrected to March 4, 1992.

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Westinghouse Non-Proprietary Class 3 C-14 Table C-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y Radially Corrected Average Measured Corrected Reaction Average Monitor Reaction Reaction Activity(a) Saturated Rate Reaction Identification Rate (Bq/g) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu - Top Mid Wire 63Cu(n,) 60Co 1.09E+05 2.25E+05 3.44E-17 Cu - Mid Wire 63Cu(n,) 60Co 1.11E+05 2.30E+05 3.50E-17 3.51E-17 3.51E-17 Cu - Bot Mid Wire 63Cu(n,) 60Co 1.14E+05 2.36E+05 3.60E-17 Fe - Top Wire 54Fe(n,p) 54Mn 9.23E+05 2.07E+06 3.28E-15 Fe - Top Mid Wire 54Fe(n,p) 54Mn 9.33E+05 2.09E+06 3.32E-15 Fe - Mid Wire 54Fe(n,p) 54Mn 9.44E+05 2.11E+06 3.35E-15 3.36E-15 3.36E-15 Fe - Bot Mid Wire 54Fe(n,p) 54Mn 9.66E+05 2.16E+06 3.43E-15 Fe - Bot Wire 54Fe(n,p) 54Mn 9.56E+05 2.14E+06 3.40E-15 Ni - Top Mid Wire 58Ni(n,p) 58Co 4.12E+06 3.40E+07 4.87E-15 Ni - Mid Wire 58Ni(n,p) 58Co 4.07E+06 3.36E+07 4.81E-15 4.83E-15 4.83E-15 Ni - Bot Mid Wire 58Ni(n,p) 58Co 4.05E+06 3.35E+07 4.79E-15 U-238 - Mid 238U(Cd)(n,f) 137Cs 6.47E+05 3.30E+06 2.16E-14 2.16E-14 1.69E-14 Np-237 - Mid 237Np(Cd)(n,f) 137Cs 4.71E+06 2.40E+07 1.51E-13 1.51E-13 1.48E-13 Co-Al - Top 59Co(n,) 60Co 1.59E+07 3.35E+07 2.19E-12 2.17E-12 2.17E-12 Co-Al - Bottom 59Co(n,) 60Co 1.57E+07 3.31E+07 2.16E-12 Co-Al - Top (Cd) 59Co(Cd)(n,) 60Co 6.86E+06 1.72E+07 1.12E-12 Co-Al - Top (Cd) 59Co(Cd)(n,) 60Co 6.88E+06 1.72E+07 1.12E-12 1.09E-12 1.09E-12 Co-Al - Bottom (Cd) 59Co(Cd)(n,) 60Co 6.51E+06 1.63E+07 1.06E-12 Co-Al - Bottom (Cd) 59Co(Cd)(n,) 60Co 6.57E+06 1.64E+07 1.07E-12 Note:

(a) Measured activities are decay corrected to March 26, 2001.

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Westinghouse Non-Proprietary Class 3 C-15 Table C-7 Comparison of Measured and Calculated Threshold Foil Reaction Rates for the In-Vessel Capsules Capsule Average Std. Dev.

Reaction T U X Y 63 Cu(n,) 60Co 1.10 1.08 1.07 1.03 1.07 2.8%

54 54 Fe(n,p) Mn 1.03 0.94 0.94 0.90 0.95 5.8%

58 Ni(n,p) 58Co 1.03 0.92 0.95 0.94 0.96 5.0%

238 137 U(n,f) Cs 1.03 1.06 0.97 0.91 0.99 6.7%

237 Np(n,f) 137Cs 1.12 1.11 1.05 1.01 1.07 4.8%

Average of M/C Results 1.01 7.0%

Table C-8 Comparison of Calculated and Best-Estimate Exposure Rates for the In-Vessel Capsules Fast (E > 1.0 MeV) Fluence Rate Iron Atom Displacement Rate Capsule BE/C Std. Dev. BE/C Std. Dev.

T 1.03 6.0% 1.04 7.0%

U 0.98 6.0% 1.00 7.0%

X 0.96 6.0% 0.97 7.0%

Y 0.92 6.0% 0.94 7.0%

Average 0.97 4.7% 0.99 4.3%

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Westinghouse Non-Proprietary Class 3 C-16 C.8 REFERENCES C-1 Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, March 2001.

C-2 Westinghouse Report WCAP-10492, Rev. 0, Analysis of Capsule T from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, March 1984.

C-3 Westinghouse Report WCAP-11554, Rev. 0, Analysis of Capsule U from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, September 1987.

C-4 Westinghouse Report WCAP-13366, Rev. 0, Analysis of Capsule X from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, June 1992.

C-5 Westinghouse Report WCAP-15692, Rev. 0, Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, August 2001.

C-6 A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

C-7 Radiation Safety Information Computational Center (RSICC) Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.

C-8 ASTM Standard E1018-09, Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (IIB), American Society for Testing and Materials, 2013.

C-9 ASTM Standard E944-13, Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIIA), American Society for Testing and Materials, 2013.

C-10 Nuclides and Isotopes: Chart of the Nuclides, 17th Edition, Lockheed Martin, 2010.

C-11 ASTM Standard E1005-16, Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, American Society for Testing and Materials, 2016.

WCAP-18571-NP July 2022 Revision 2

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LR-N22-0012 LAR S22-01 Enclosure 5 Westinghouse Letter Report LTR-SCS-20-52-NP, Rev. 1, Salem Unit 2 Pressurizer Overpressure Protection System (POPS) Analysis, (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 To: Tim L. OConnor Date: July 26, 2021 From: Functional, Systems & Setpoints Engineering Phone: 412-374-2779 Our ref: LTR-SCS-20-52-NP, Rev. 1

Subject:

Salem Unit 2 Low Temperature Overpressure Protection System (LTOPS) / Pressurizer Overpressure Protection System (POPS) Analysis

Reference:

1. LTR-AMER-MKG-18-1118, Revision 1, Westinghouse Revised Offer for Pressure-Temperature Limits Support and Surveillance Capsule Removal, Testing and Analysis for Salem Units 1 and 2, October 2018.

2. PSE-09-18, Revision 0, Transmittal of LTOPS Setpoint Evaluation for Salem Units 1 and 2, March 2009.

This letter transmits the Pressurizer Overpressure Protection System (POPS) analysis report for Salem Unit 2 to Public Service Electric and Gas (PSEG) Nuclear. Per the Reference 1 offer, this analysis was performed to validate the applicability of the setpoints in the previous POPS analysis performed in Reference 2. This evaluation confirms that the current POPS pressurizer Power Operated Relief Valve (PORV) settings remain valid. The Westinghouse Non-Proprietary Class 3 version of the analysis is provided in Attachment 1 with the proprietary information identified within brackets removed.

The attachment to this letter will be transmitted to PSEG along with an application for withholding proprietary information from public disclosure and supporting affidavit. The types of proprietary information are identified via superscripts following each bracket, which correspond to the types described in item 5 of the corresponding application for withholding proprietary information.

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Westinghouse Non-Proprietary Class 3 LTR-SCS-20-52-NP, Rev. 1 Page 2 of 2 Please contact the undersigned with any questions.

Luke J. Mitchell*

Functional, Systems & Setpoints Engineering Verifier: Thomas G. Joseph* Manager: Steven R. Billman*

Functional, Systems & Setpoints Engineering Functional, Systems & Setpoints Engineering Attachment 1: Salem Unit 2 Pressurizer Overpressure Protection System (POPS) Analysis (Non-Proprietary)

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

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Westinghouse Non-Proprietary Class 3 Page 1 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 Salem Unit 2 Pressurizer Overpressure Protection System (POPS) Analysis July 2021 (9 Pages)

Luke J. Mitchell Functional, Systems & Setpoints Engineering Thomas G. Joseph Functional, Systems & Setpoints Engineering

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 2 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 1.0 Introduction The Low Temperature Overpressure Protection System (LTOPS), known as the Pressurizer Overpressure Protection System (POPS) at Salem, provides Reactor Coolant System (RCS) pressure relief capability to mitigate the overpressure transients that may occur during cold shutdown, heatup, and cooldown operations to minimize the potential for challenging reactor vessel integrity limits when operating at low temperature conditions (i.e., 10 CFR 50, Appendix G limits). At Salem, the pressurizer Power Operated Relief Valves (PORVs), with reduced lift settings, provide a method of POP. The POPS PORV setpoints are selected in accordance with the NRC approved methodology in Reference 1 such that the peak pressure during the design basis Mass Injection (MI) and Heat Injection (HI) transients will not exceed the isothermal Appendix G Pressure-Temperature (P-T) limits.

This evaluation will demonstrate that the current POPS analysis documented in Reference 2 remains valid and any changes to the 50 Effective Full-Power Year (EFPY) P-T limit curve applicability term will be reconciled. The current POPS analysis was performed for the 50 EFPY P-T limit curves documented in Reference 3. The 50 EFPY P-T limit curve applicability term was subsequently confirmed to remain valid in Reference 4.

The scope of this evaluation will include reviewing inputs from Reference 1 to determine if any have changed after those analyses were completed. Any changes will be evaluated to determine if POPS PORV settings need to be updated. This work was completed in accordance with the scope defined in Reference 5.

1.1 Limits of Applicability The results of this evaluation are applicable to Salem Unit 2 with the key input parameters defined in Section 2.1 and the Reference 3 P-T limits, which were shown to remain valid through 50 EFPY in Reference 4.

1.2 Open Items None.

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 3 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 2.0 Input Parameters and Assumptions 2.1 Key Inputs The key analysis input parameters used in the development of the Salem Unit 2 POPS setpoints are summarized in the list below. A detailed list of the inputs was sent to PSEG in Reference 6 for confirmation. The confirmation of the key POPS input parameters was received in Reference 7.

1. Design Basis Mass Injection (MI) Transient The design basis MI event is defined as flow into a water-solid RCS as a result of either, (a) one High Head Safety Injection (HHSI) pump and one Positive Displacement Pump (PDP), or (b) one Intermediate Head Safety Injection (IHSI) pump; whichever is greater. For the analysis performed in Reference 2, a constant IHSI pump flow rate of [ ]a,c was specified as the bounding MI flow rate. PSEG confirmed this value in Reference 7.

2. Design Basis Heat Injection (HI) Transient The HI transient is defined as the inadvertent startup of one RCP with the SG secondary side a maximum of 50°F hotter than each of the RCS cold leg temperatures. Prior to the RCP start, all loops are inactive and the entire RCS primary side (except for stagnant water in the SG tubes) is assumed to be 50°F cooler than the secondary side. PSEG confirmed this in Reference 7.

3. Pressurizer PORV Stroke and Delay Times The PORV characteristics from the analysis of record (Reference 2), confirmed in Reference 7, are as follows.

Valve type = Copes-Vulcan Full open CV (for sub-cooled water discharge) = [ ]a,c gpm / psi Opening Time = [ ] a,c sec (full close to full open)

Closing Time = [ ] a,c sec (full open to full close)

Delay Time (open and close) = [ ] a,c sec1 Valve curve (CV vs. lift) = See Figure 1 Maximum pressure relief tank backpressure2 = [ ] a,c psig PORV Reset Pressure = FSAR Section 7.6.3.2 implies the valve closes when pressure decreases below the opening pressure (currently 375 psig), [

] a,c.

1 Reference 7 confirmed a signal delay of [ ] a,c seconds. This may not include valve solenoid and seat delays and is bounded by the previously modeled Reference 2 value of [ ] a,c seconds. Therefore, the current delay time of [ ] a,c seconds is maintained for this analysis.

2 The backpressure is used to calculate the PORV relief capacity as a function of pressurizer pressure. A higher backpressure results in a lower effective PORV capacity and is limiting for the analysis.

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 5 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1

6. RCP Operating Restrictions Per Reference 7, Salem Unit 2 currently utilizes the following restrictions on RCP operation:

No more than two RCPs may be operating at RCS Temperatures (TRCS) 200°F All four RCPs are allowed to operate at TRCS > 200°F.

7. PORV Piping Limit In addition to the Appendix G limits, an 800 psig pressure limit is included to address pressurizer PORV piping load conditions associated with subcooled water discharge (confirmed in Reference 7). This limit is recognized as an operational consideration that is accommodated by the POPS in Reference 1. The PORV piping has been generically evaluated for the water hammer loads associated with cyclic water relief at up to 800 psig. Therefore, when the plant is operated water solid, the POPS settings ensure that the pressure does not exceed the design value of 800 psig.

8. Reactor Coolant Pump (RCP) No. 1 Seal P Limits As discussed in Reference 1, the POPS analysis generally evaluates the minimum RCS pressure following POPS actuation compared to the minimum RCS pressure for RCP operation. This represents an operational consideration intended to maintain the health of the RCP #1 seals. When there is a conflict between meeting the upper limit (i.e., Appendix G P-T limits) and lower limit (i.e., RCP seal limit), the upper limit takes precedence. [

] a,c Therefore, consistent with the current Reference 2 analysis, the RCP seal P limit will not be explicitly evaluated. [

] a,c

9. Current POPS PORV Settings:

The current analysis in Reference 2 developed a single POPS PORV setpoint of 434 psig for 50 EFPY. PSEG confirmed in Reference 7 that Salem Unit 2 will continue to use a single POPS PORV setting value. It should be noted that the current POPS PORV setpoint of 375 psig at Salem Unit 2 remains conservative with respect to the setting developed in Reference 2.

10. Current POPS Enable Temperature:

The current POPS enable temperature documented in the Salem Unit 2 Technical Specifications is 312°F. The minimum EOLE POPS enable temperature defined in Reference 3 is 292°F (includes uncertainty), which was revalidated through 50 EFPY in Reference 4. Therefore, the current enable temperature defined in the Technical Specifications bounds the minimum EOLE POPS enable temperature defined in Reference 4. In addition, the current POPS enable temperature for Salem Unit 1 was calculated to be [ ] a,c °F in Reference 8. If desired for consistency between both units, PSEG can choose to implement the [ ] a,c °F enable temperature for Salem Unit 2 since it bounds both the current enable temperature documented in the Technical Specifications and the minimum EOLE POPS enable temperature defined in Reference 4.

11. Appendix G P-T Limits The current POPS analysis was performed for the 50 EFPY P-T limit curves documented in Reference 3. The 50 EFPY P-T limit curves developed in Reference 3 were subsequently confirmed to remain valid for implementation in Reference 4.

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 6 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 In accordance with the Reference 1 methodology, the POPS must protect the steady state (isothermal) P-T limits. Therefore, Table 1 summarizes the current isothermal P-T limits used in the calculation of the POPS PORV settings, which remain valid to 50 EFPY.

Table 1: Steady State Appendix G Limits Valid through 50 EFPY for Salem Unit 2 (without Uncertainties)

Salem Unit 2 for 50 EFPY RCS Temperature (°F) Appendix G limit RCS Temperature (°F) Appendix G limit (psig)

(psig) 60 621 175 799 65 621 180 819 70 621 185 842 75 621 190 866 80 621 195 894 85 621 200 924 90 621 205 957 95 621 210 994 100 621 215 1034 105 621 220 1079 110 621 225 1129 115 621 230 1184 120 621 235 1245 125 621 240 1312 130 621 245 1386 135 621 250 1468 135.1 693 255 1558 140 702 260 1658 145 712 265 1769 150 723 270 1891 155 736 275 2026 160 749 280 2176 165 764 285 2341 170 781

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 7 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 2.2 Key Assumptions The following assumptions are applicable for the Salem Unit 2 POPS setpoint analysis. Unless otherwise noted, these assumptions are consistent with the current analysis in Reference 2.

1. It is assumed that the RCS is enclosed by a non-yielding, inelastic boundary. The pressurizer is assumed to be in a water solid condition with the water at the same subcooled temperature as the remainder of the RCS. [

] a,c

2. Only one PORV is credited to mitigate the low temperature overpressure event to meet the single failure criteria.

3. All MI cases were analyzed at an RCS temperature of 60°F, which is the minimum RCS temperature corresponding to the bolt up temperature in Reference 3. [

] a,c

4. For the HI transient the entire RCS primary side, with the exception of the water in the SG tubes, is conservatively assumed to be 50°F cooler than the SG secondary side temperature in all four SGs.

5. A single-phase, sub-cooled water discharge through the PORV was assumed.

6. Letdown flow is conservatively assumed to be isolated during the MI and HI transients.

[

] a,c

7. [

] a,c 3.0 Description of Evaluation This evaluation is limited in scope, since the purpose is to ensure that no inputs to the Salem Unit 2 POPS analysis have changed since the current analysis in Reference 2 was performed. This evaluation will review the inputs used in the current analysis and ensure that each input is still valid. Therefore, the methodology used for this evaluation will first consist of performing a cursory review of the current analysis in Reference 2. This review will identify all key inputs and how they are applied in developing the current POPS PORV setting. Then, these inputs will be validated against the current design inputs that were confirmed in Reference 7. If the input parameters have changed since the base analyses, the resulting impact will be reviewed to determine if the POPS analysis will need to be revised.

An analysis was performed in Reference 9 to support the Technical Specification LCO 3.4.1.3 and 3.4.1.4 start notes which suggest that an RCP can be started if the pressurizer water volume is less than 1650 ft3 (approximately 92%) with no limitation on the magnitude of a potential secondary-to-primary temperature asymmetry. The analysis in Reference 9 showed that the analyses performed in Reference 2 with water solid conditions remain limiting for the calculation of the POPS PORV setting and that the LCO 3.4.1.3 and 3.4.1.4 start notes are acceptable.

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 8 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 4.0 Acceptance Criteria The following acceptance criteria are used to determine the POPS PORV setpoint for Salem Unit 2 per Reference 2, and are reiterated here for this evaluation:

1. The peak RCS pressure resulting from the design basis MI and HI transients shall not exceed the lower of the following for all applicable temperatures:

o Maximum allowable pressure of the steady-state 10 CFR 50, Appendix G reactor vessel P-T limits o 800 psig (PORV inlet pressure discharge piping limit)

2. The minimum RCS pressure resulting from the design basis MI and HI transients should not drop below the RCP No. 1 Seal P limit.

If there is a conflict between satisfying the upper limits (i.e., the minimum of the Appendix G limits and the piping limit) and the lower limits (i.e., the RCP No. 1 Seal P limits), the upper pressure limits will take precedence. As noted in item #8 of Section 2.1, criterion 2 is not explicitly evaluated [

] a,c.

5.0 Results and Conclusions Based on the confirmation of key POPS input parameters from PSEG in Reference 7, it is concluded that no input parameters from the current analysis in Reference 2 have been changed. Reference 4 concluded that the 50 EFPY P-T limits in Reference 3 remain valid through 50 EFPY. Since none of the key POPS input parameters have changed, and the NRC approved methodology in Reference 1 also has not changed, it is concluded that the acceptance criteria outlined in Section 4.0 of this evaluation continue to be met for the POPS settings determined in Reference 2. Specifically, a single pressurizer PORV (both PORVs are required to be operable to address a single failure), with the maximum allowable POPS pressurizer PORV setting of 434 psig, is capable of providing protection across the full temperature range applicable to POP. The setting requires that no more than two Reactor Coolant Pumps (RCPs) be in operation at indicated RCS temperatures 200°F. This is consistent with current RCP operating restrictions provided in Reference 7.

Since the POPS PORV setting calculated for 50 EFPY remains bounded by the current technical specification setting of 375 psig, PSEG can either choose to maintain the current 375 psig setting or increase the setting to the current maximum allowable setpoint of 434 psig. Increasing the POPS PORV setting would increase operating margin to POPS actuation during heatup and cooldown operations.

The current POPS enable temperature of 312°F documented in the Salem Unit 2 Technical Specifications bounds the minimum EOLE POPS enable temperature of 292°F (includes uncertainty) defined in Reference 3, which was revalidated through 50 EFPY in Reference 4. In addition, the current POPS enable temperature for Salem Unit 1 was calculated to be [ ] a,c°F in Reference 8. The minimum calculated enable temperature for 50 EFPY is less restrictive than both the current Unit 2 value and the revised Unit 1 value. PSEG can therefore implement a POPS enable temperature as low as 292°F, maintain the current technical specification value of 312°F, or use the Unit 1 value of [ ] a,c °F for consistency between units.

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 9 of 9 LTR-SCS-20-52-NP, Rev. 1 Attachment 1 6.0 References

1. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

2. a,c

3. WCAP-16982-NP, Revision 2, Salem Unit 2 Time-Limited Aging Analysis on Reactor Vessel Integrity, July 2020.

4. WCAP-18571-NP, Revision 0, Verification of the Salem Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, September 2020.

5. a,c 6.

7.

8. LTR-SCS-20-28-NP, Revision 1, Salem Unit 1 Low Temperature Overpressure Protection System (LTOPS) / Pressurizer Overpressure Protection System (POPS) Analysis, February 2021.

a,c 9.

- - This record was final approved on 7/28/2021 11:40:33 AM. (This statement was added by the PRIME system upon its validation)

LR-N22-0012 LAR S22-01 Enclosure 6 Affidavit for Withholding

Westinghouse Non-Proprietary Class 3 CAW-21-5205 Page 1 of 3 COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1) I, Anthony J. Schoedel, 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 LTR-SCS-20-52-P, Revision 1 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

Westinghouse Non-Proprietary Class 3 CAW-21-5205 Page 2 of 3 others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

(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:

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v t e Letter Sequence Request
EPID:L-2022-LLA-0111, TSTF-363, TSTF-408, TSTF-419, TSTF-233 (Approved, Closed)
Initiation Request Acceptance Administration Withholding Request Acceptance Questions Answers Results Approval Other:
MONTHYEAR2022-09-01 [Table View]

LR-N22-0012, License Amendment Request to Amend the Technical Specifications to Revise and Relocate the Reactor Coolant System Pressure and Temperature Limits and Pressurizer Overpressure Protection System Limits to a Pressure and Temperature

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SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

3.1 Background

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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References:

Subject:

REFERENCES:

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SUMMARY

SUMMARY

SUMMARY

2.1 INTRODUCTION

SUMMARY

2.1 INTRODUCTION

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