Text

Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
	ML22343A257
Person / Time
Site:	Prairie Island
Issue date:	12/02/2022
From:	Domingos C; Northern States Power Co, Xcel Energy
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML22343A256	List: L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR);
References
	L-PI-22-020
	Download: ML22343A257 (1)
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ENCLOSURE 5 CONTAINS PROPRIETARY INFORMATION WITHHOLD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 1717 Wakonade Drive Welch, MN 55089 December 2, 2022 L-PI-22-020 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR)

Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), hereby submits a request for an amendment to the Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and Unit 2.

The proposed amendment would revise PINGP TS 5.6.6, Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR), to replace the current PTLR method with more recent analytical methods that have been found acceptable by the NRC. The proposed amendment also removes from TS 5.6.6 a reference to an American Society of Mechanical Engineers (ASME) Code Case that is no longer necessary.

provides a description and assessment of the proposed change. Attachment 1 to provides the existing TS pages marked to show the proposed changes. to Enclosure 1 provides revised (clean) TS pages.

provides a copy of non-proprietary topical report WCAP-18746-NP, Revision 1, Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, which was prepared using the proposed methods.

provides a redacted (non-proprietary) version of Enclosure 5 and Enclosure 4 is an affidavit from Westinghouse in support of withholding Enclosure 5 from public disclosure.

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Xcel Energy

Document Control Desk L-Pl-22-020 Page 2 provides a copy of proprietary Westinghouse calculation NSPM-L TP-TR-AA-000001, Revision 0, "Prairie Island Units 1 and 2 Low Temperature Overpressure Protection System (LTOPS) Analysis," which was prepared using the proposed methods.

NSPM requests approval of the proposed amendment 12 months after NRC acceptance.

Once approved, the amendment shall be implemented within 90 days.

In accordance with 10 CFR 50.91, a copy of this application, excluding the proprietary enclosure, is being provided to the designated Minnesota Official.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

Please contact Mr. Jeff Kivi at (612) 330-5788 or Jeffrey.L.Kivi@xcelenergy.com if there are any questions or if additional information is needed.

I declare under penalty of perjury, that the foregoing is true and correct. Executed on li.lt.

, 2022.

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Christopher P. Domingos Site Vice President, Monticello and Prairie Island Nuclear Generating Plants Northern States Power Company - Minnesota Enclosures cc:

Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC State of Minnesota

Page 1 of 13 ENCLOSURE 1 PRAIRE ISLAND NUCLEAR GENERATING PLANT License Amendment Request:

Revise Technical Specification 5.6.6, Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR) 1.0

SUMMARY

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

3.0 TECHNICAL EVALUATION

............................................................................................. 4 3.1 WCAP-14040-A Methodology...................................................................................... 4 3.2 WCAP-18124-NP-A and Supplement 1-NP-A Methodology......................................... 6 3.3 Conclusion.................................................................................................................... 7

4.0 REGULATORY EVALUATION

........................................................................................ 7 4.1 Applicable Regulatory Requirements/Criteria............................................................... 7 4.1.1 10 CFR Part 50 and NRC Regulatory Guides........................................................ 7 4.1.2 Principle Design Criteria (PDC).............................................................................. 9 4.2 Precedent................................................................................................................... 10 4.3 No Significant Hazards Consideration Analysis.......................................................... 10 4.4 Conclusions................................................................................................................ 12

5.0 ENVIRONMENTAL CONSIDERATION

......................................................................... 12

6.0 REFERENCES

.............................................................................................................. 12 ATTACHMENTS:

1. Proposed Technical Specification Changes (Mark-up)

2. Proposed Technical Specification Changes (Clean)

L-PI-22-020 NSPM Page 2 of 13 DESCRIPTION AND ASSESSMENT 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), hereby submits a request for an amendment to the Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and Unit 2.

The proposed amendment would revise PINGP TS 5.6.6, Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR), to replace the current PTLR method with more recent analytical methods that have been found acceptable by the NRC. The proposed amendment also removes from TS 5.6.6 a reference to American Society of Mechanical Engineers (ASME) Code Case N-514, which is no longer necessary.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Components of the RCS are designed to withstand effects of loadings due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. Pressure and temperature changes are limited during RCS heatup and cooldown within the design assumptions and the stress limits for normal operation.

Pressure and temperature (PT) limits have been established for heatup, cooldown, inservice leak and hydrostatic testing and are documented in the PINGP PTLR.

2.2 Current Technical Specification Requirements Currently, PINGP TS Section 5.6.6, part b, states:

The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves (includes any exemption granted by NRC to ASME Code Case N-514).

2.3 Reason for Proposed Change NSPM is proposing to change the analytical method used to determine the RCS PT limits and cold overpressure mitigating system (COMS) setpoints. Additionally, NSPM proposes a

L-PI-22-020 NSPM Page 3 of 13 change to the analytical method required for calculating neutron fluence. Finally, the change will also allow removal of reference to ASME Code Case N-514.

The proposed change updates the approved analytical method to determine RCS PT limits and COMS setpoints to the current Westinghouse methodology provided in WCAP-14040-NP-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves. The NRC approved WCAP-14040-NP-A, Revision 4, for use by safety evaluation (SE) dated February 27, 2004 (Reference 1).

The proposed change updates the approved analytical method for calculating neutron fluence to the current Westinghouse methodology provided in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Fluence Determination with RAPTOR-M3G and FERRET -

Supplement for Extended Beltline Materials. These methodologies would be used in lieu of the method provided in WCAP-14040-NP-A and adhere to the guidance in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The NRC approved WCAP-18124-NP-A, Revision 0, for use by SE dated June 15, 2018 (Reference 5) and approved WCAP-18124-NP-A, Revision 0,Supplement 1-NP-A, for use by SE dated April 20, 2022 (Reference 7).

The proposed change removes discussion of ASME Code Case N-514. The NRC approved an exemption for PINGP by letter dated April 30, 1998, which allowed NSPM to use Code Case N-514 to determine each units COMS setpoint (Reference 2). ASME has subsequently incorporated Code Case N-514 into later editions of the ASME Code approved by the NRC.

Under the proposed new analytical method of WCAP-14040-NP-A, Revision 4, NSPM will no longer need to apply the exemption to apply ASME Code Case N-514, which ASME annulled in 2002.

2.4 Description of Proposed Change TS 5.6.6.b.

Replace:

The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves (includes any exemption granted by NRC to ASME Code Case N-514).

With:

The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigating System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

L-PI-22-020 NSPM Page 4 of 13

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

2. WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with Raptor-M3G and FERRET -

Supplement for Extended Beltline Materials, May 2022, shall be used as an alternative to Section 2.2 of WCAP-14040-NP-A.

3.0 TECHNICAL EVALUATION

3.1 WCAP-14040-A Methodology The analytical method to be used to determine the revised RCS PT limits and COMS setpoints for PINGP is described in the NRC-approved Topical Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

NSPM has reviewed WCAP-14040-NP-A, Revision 4, and determined that it is applicable to this facility, and that use of the new methodology will provide the appropriate operating curves and limits for operation to the end of the renewed facility operating licenses at the licensed power level. When required by 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, revised PINGP PT limits will be developed using the NRC-approved Topical Report WCAP-14040-A, Revision 4. COMS setpoints will also be developed using the aforementioned methodology. Since the methodology is approved and accepted for use by the NRC, it provides an acceptable means of satisfying 10 CFR 50, Appendix G, Fracture Toughness Requirements, which governs the development of PT limits and COMS setpoints.

The NRC has found the methodology contained in topical report WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 3, acceptable for referencing in licensing applications subject to three conditions as described in the February 27, 2004, NRC final safety evaluation. Revision 3 of the topical report was reissued in May of 2004 as Revision 4 (Reference 1) with the referenced NRC acceptance letter and safety evaluation incorporated.

The three conditions and responses indicating how each condition has been addressed are presented below.

1. 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 Generic Letter 96-03 related to the reactor pressure vessel material surveillance program.

L-PI-22-020 NSPM Page 5 of 13

Response

The minimum methodology requirements associated with Provision 2 in the table of to Generic Letter 96-03 require that licensees:

Briefly describe the surveillance program. Licensees should identify by title and number report containing the Reactor Vessel Surveillance Program and surveillance capsule reports.

The PINGP PTLR describes the methodology requirements discussed in Provision 2 in the table of Attachment 1 to Generic Letter 96-03 related to the reactor pressure vessel material surveillance program. The program complies with Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements." The PINGP PTLR also includes the latest surveillance specimen removal schedules.

The most recent post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens for PINGP Unit 1 was performed in accordance with 10 CFR 50, Appendix H and American Society for Testing and Materials (ASTM) Specification E185-82 (Reference 9). The most recent post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens for PINGP Unit 2 was performed in accordance with 10 CFR 50, Appendix H and ASTM Specification E185-82 (Report in progress).

The NRC approved the current Reactor Vessel Surveillance Capsule Withdrawal Schedule for PINGP Units 1 and 2 in Reference 8.

2. Contrary to the information in WCAP-14040, Revision 3, licensees use of the provisions of ASME Code Cases N-588, N-640, or N-641 in conjunction with the basic methodology contained 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 topical report WCAP-14040 should be modified to reflect this NRC staff conclusion.

Response

Table A-1, "Status of ASME Nuclear Code Cases Associated with the P-T Limit Curve/COMS Methodology," of Appendix A, "Relevant ASME Nuclear Code Cases" was updated in Revision 4 of WCAP-14040-A to indicate the date, edition, and addenda of ASME Code,Section XI, when the referenced code cases were approved by the ASME. As indicated in Condition 2, the editions and addenda of ASME Code,Section XI, that are listed in Table A-1 have been incorporated by reference in 10 CFR 50.55a.

3. As stated in WCAP-14040, Revision 3, until Appendix G to 10 CFR Part 50 is revised to modify or eliminate the existing reactor pressure vessel flange minimum temperature requirements or an exemption request to modify or eliminate these requirements is

L-PI-22-020 NSPM Page 6 of 13 approved by the NRC for a specific facility, the stated minimum temperature must be incorporated into a facility's PT limit curves.

Response

This amendment proposes that Revision 4 of WCAP-14040-A be implemented for the PTLR methodology at PINGP. In accordance with Section 2.9 of WCAP-14040-A, Revision 4, the reactor pressure vessel flange minimum temperature requirement will continue to be incorporated into the PINGP PT limit curves, until Appendix G to 10 CFR Part 50 is revised to modify or eliminate the existing reactor pressure vessel flange requirements, or an exemption request to modify or eliminate these requirements is approved by the NRC.

TS 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),"

requires revision to reflect the use of the above-mentioned alternative methodology.

Additionally, revision is needed to remove reference to ASME Code Case N-514, which has been annulled after being incorporated in a subsequent revision of the ASME Code.

3.2 WCAP-18124-NP-A and Supplement 1-NP-A Methodology The NRC staff has found the calculational fluence methodology described in WCAP-18124-NP-A, Revision 0, acceptable for use in calculating reactor pressure vessel neutron fluence subject to two conditions described in the NRC final safety evaluation (Reference 5). The two conditions and responses indicating how the conditions will be addressed are presented below.

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 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.

Response

Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, (Reference 6) states that any materials predicted to exceed 1.0 x 1017 n/cm2 (E > 1.0 MeV) at the end of their licensed operating period must be evaluated to determine the changes in fracture toughness. Materials that are not adjacent to the active core yet are predicted to accrue fluence levels greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) are now commonly referred to as extended beltline materials.

WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, provided the

L-PI-22-020 NSPM Page 7 of 13 justification necessary to narrow Limitation #1 and allow licensees to apply the RAPTOR-M3G method in the extended beltline regions of RPVs on a generic basis. The NRC staff has determined that the fluence methods and qualifications described in WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A are acceptable for referencing in licensing applications as described in the NRC final safety evaluation (Reference 7).

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.

Response

For PINGP, least squares adjustment analysis is not used to modify the calculated surveillance capsule or reactor pressure vessel neutron exposure when RAPTOR-M3G methodology is applied. If least squares adjustment is used to modify calculated fluence values in future reactor vessel integrity evaluations, the requirements of Limitation 2 will be met.

3.3 Conclusion The proposed change in methodology for determining reactor coolant system PT limits and COMS setpoints, and for calculating neutron fluence values is acceptable since the conditions and limitations identified by the NRC staff would be addressed as described above, and the methodologies have previously been found acceptable by the staff.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.1.1 10 CFR Part 50 and NRC Regulatory Guides The NRC has established requirements in 10 CFR Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The acceptability of a facility's proposed PTLR methodology is based on the NRC regulations and guidance as discussed below.

10 CFR 50.36(c)(5), "Administrative controls," are the provisions relating to organization and management, procedures, recordkeeping, review, and audit, and reporting necessary to assure operation of the facility in a safe manner. The proposed amendment would continue to assure operation of the facility in a safe manner by updating the methodology used for determining reactor coolant system PT limits and COMS setpoints, and for calculating neutron fluence values specified in the administrative controls section of Technical Specifications to more recent methodology found acceptable by the NRC staff.

L-PI-22-020 NSPM Page 8 of 13 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for light water nuclear power reactors for normal operation," paragraph (a) states:

Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the certifications required under

§50.82(a)(1) have been submitted, must meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in appendices G and H to this part.

10 CFR 50, Appendix G, "Fracture Toughness Requirements," requires the establishment of PT limits for specific material fracture toughness requirements of the reactor coolant pressure boundary materials. 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," establishes requirements related to facility reactor pressure vessel material surveillance programs.

In Reference 1, the NRC staff concluded that the methodology specified in WCAP-14040-NP-A, Revision 4, addresses reactor pressure vessel minimum temperature requirements in a way which is consistent with 10 CFR 50, Appendix G, and ASME Code,Section XI, Appendix G. The NRC Staff also concluded that the basic methodology specified in WCAP-14040-NP-A, Revision 4, for establishing PT limits meets the regulatory requirements of 10 CFR 50, Appendix G, and the guidance provided in Standard Review Plan, Section 5.3.2, "Pressure-Temperature Limits."

The PINGP Reactor Vessel Material Surveillance programs comply with the requirements of 10 CFR 50, Appendix H. Implementation of the proposed amendment will not affect the compliance of the Reactor Vessel Material Surveillance program with 10 CFR 50, Appendix H.

To satisfy the requirements of both 10 CFR 50, Appendix G, and 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events," methods for determining the fast neutron fluence are necessary to estimate the fracture toughness of the pressure vessel materials. Regulatory Guide 1.190, "Calculational and Dosimetry Methods For Determining Pressure Vessel Neutron Fluence," provides state-of-the-art calculations and measurement procedures that are acceptable to the NRC staff for determining pressure vessel fluence.

In Reference 5, the NRC staff reviewed WCAP-18124-NP-A Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," and determined that the proposed methodology adheres to the guidance of Regulatory Guide 1.190. The NRC concluded that WCAP-18124-NP-A is acceptable for use in calculating reactor pressure vessel neutron fluence provided that the limitations and conditions listed in Section 4.0 of the NRC safety evaluation report were met. In Reference 7, the NRC staff reviewed WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, and determined that appropriate modeling techniques and adequate qualification were provided to apply RAPTOR-M3G to determine neutron fluence in the reactor vessel extended beltline and that the modeling techniques

L-PI-22-020 NSPM Page 9 of 13 adhere to the guidance in Regulatory Guide 1.190, as appropriate, and exceed it when necessary.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

describes procedures acceptable to the NRC staff for calculating the effects of neutron irradiation embrittlement of the low-alloy steels of reactor pressure vessels. In Reference 1, the NRC staff determined that the methodology described for determining material adjusted reference temperature values in WCAP-14040, Revision 3 was consistent with the guidance provided in the ASME Code, Standard Review Plan Section 5.3.1, "Reactor Vessel Materials,"

and Regulatory Guide 1.99, Revision 2, and was, therefore, acceptable.

Based on the foregoing, the proposed amendment will continue to ensure compliance with the above referenced regulations or guidance and will ensure that the functional capabilities or performance levels of equipment required for safe operation are met.

4.1.2 Principle Design Criteria (PDC)

PINGP was not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC) and was designed and constructed to comply with NSPMs understanding of the intent of the Atomic Energy Commission General Design Criteria for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. Since the construction of the plant was significantly completed prior to the issuance of the February 20, 1971, 10CFR50, Appendix A GDC, the plant was not reanalyzed and the Final Safety Analysis Report (FSAR) was not revised to reflect these later criteria. However, the AEC Safety Evaluation Report acknowledged that the AEC staff assessed the plant, as described in the FSAR, against the Appendix A design criteria and... are satisfied that the plant design generally conforms to the intent of these criteria. The following discussion addresses the proposed changes with respect to meeting the requirements of the applicable draft design criteria to which PINGP Unit 1 and Unit 2 are licensed.

GDC 9 - Reactor Coolant Pressure Boundary: The reactor coolant pressure boundary shall be designed, fabricated, and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime.

GDC 34 - Pressure Boundary Rapid Propagation Failure Prevention: The reactor coolant pressure boundary shall be designed and operated to reduce to an acceptable level the probability of rapidly propagating type failures. Consideration is given (a) to the provisions for control over service temperature and irradiation effects which may require operational restrictions, (b) to the design and construction of the reactor pressure vessel in accordance with applicable codes, including those which establish requirements for absorption of energy within the elastic strain energy range and for absorption of energy by plastic deformation and (c) to the design and construction of reactor coolant pressure boundary piping and equipment in accordance with applicable codes.

GDC 35 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention. Under conditions where reactor coolant pressure boundary system components constructed of ferritic materials

L-PI-22-020 NSPM Page 10 of 13 may be subjected to potential loadings, such as a reactivity-induced loading, service temperature shall be at least 120 degrees F above the nil ductility transition (NDT) temperature of the component material if the resulting energy release is expected to be absorbed by plastic deformation or 60 degrees F above the NDT temperature of the component material if the resulting energy release is expected to be absorbed within the elastic strain energy range.

The proposed amendment would revise TS 5.6.6 to add the current approved Westinghouse analytical methods for determining RCS PT limits and COMS setpoints. The proposed amendment would also remove reference to annulled ASME Code Case N-514. The basic methodology of WCAP-14040-NP-A, Revision 4, and WCAP-18124-NP-A, Revision 0 (including Supplement 1-NP-A, Revision 0) for establishing PT limits meets the regulatory requirements of Appendix G to 10 CFR Part 50. Therefore, PINGP PDC will continue to be met as RPV integrity continues to be assured.

4.2 Precedent The proposed amendment would allow use of a new method, as described in Topical Report WCAP-14040-A, Revision 4, and WCAP-18124-NP-A, for the development of PT limits and COMS setpoints. The NRC previously approved the WCAP-14040-A, Revision 4 methodology for the development of PT limits and COMS setpoints at:

Beaver Valley Units 1 and 2, by a safety evaluation dated November 1, 2021 (Reference 4)

Point Beach Units 1 and 2, by a safety evaluation dated June 30, 2014 (Reference 3)

The NRC previously approved the WCAP-18124-NP-A methodology for the calculation of fast neutron fluence in ferritic components of the reactor pressure vessel as an alternative to the fluence methods described in Section 2.2 of WCAP-14040-A, Revision 4, at:

Beaver Valley Units 1 and 2, by a safety evaluation dated November 1, 2021 (Reference 4) 4.3 No Significant Hazards Consideration Analysis Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), proposes to revise PINGP TS 5.6.6, Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR), to update the approved analytical method to determine RCS pressure and temperature (PT) limits and cold overpressure mitigating system (COMS) setpoints to the current Westinghouse methodology provided in WCAP-14040-NP-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves. The proposed change also updates the approved analytical method for calculating neutron fluence to the current Westinghouse methodology provided in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials. The proposed amendment also removes from TS 5.6.6 a reference to an American Society of Mechanical Engineers (ASME) Code Case that is no longer necessary.

L-PI-22-020 NSPM Page 11 of 13 NSPM has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:

1.

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

Response: No The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or the manner in which the plant is operated and maintained. The proposed change does not alter or prevent the ability of structures, systems, or components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

There will be no adverse change to normal plant operating parameters, engineered safety feature actuation setpoints, accident mitigation capabilities, or accident analysis assumptions or inputs. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

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

2.

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

Response: No The proposed change does not impose any new or different requirements or eliminate any existing requirements. The proposed change is consistent with the current safety analysis assumptions and current plant operating practice. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. Equipment important to safety will continue to operate as designed. The change does not result in any event previously deemed incredible being made credible. The change does not result in adverse conditions or result in any increase in the challenges to safety systems.

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

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No

L-PI-22-020 NSPM Page 12 of 13 The proposed change does not alter safety limits, limiting safety system settings, or limiting conditions for operation. The setpoints at which protective actions are initiated are not altered by the proposed change. There are no new or significant changes to the initial conditions contributing to accident severity or consequences. The proposed amendment will not otherwise affect the plant protective boundaries, will not cause a release of fission products to the public, nor will it degrade the performance of any other structures, systems, or components important to safety.

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

Based on the above, NSPM concludes that the proposed amendment does not involve a 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 In conclusion, 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 Commission's 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

The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(10). 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. NRC letter to Westinghouse Electric Company, dated February 27, 2004, Final Safety Evaluation for Topical Report WCAP-14040, Revision 3, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (TAC No. MB5754). (ADAMS Accession No. ML040620297)

2. NRC letter to NSPM, dated April 30, 1998, Exemption from 10 CFR 50.60 by Applying ASME Code Case N-514 for Prairie Island Nuclear Generating Plant, Units 1 and 2 (TAC Nos. MA0682 and MA0683). (ADAMS Accession No. ML022260669)

L-PI-22-020 NSPM Page 13 of 13

3. NRC letter to NextEra Energy Point Beach, LLC, dated June 30, 2014, Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendment Regarding Change to Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (TAC Nos. MF0532 and MF0533). (ADAMS Accession No. ML14126A378)

4. NRC letter to Energy Harbor Nuclear Corp. dated November 1, 2021, Beaver Valley Power Station, Units 1 and 2 - Issuance of Amendment Nos. 313 and 203 re: Reactor Coolant System, Pressure and Temperature Limits Report (EPID L-2020-LLA-0233). (ADAMS Accession No. ML21197A009)

5. NRC letter to Westinghouse Electric Company, dated June 15, 2018, Final Safety Evaluation for Westinghouse Electric Company Topical Report WCAP-18124-NP, Revision 0, Fluence Determination with RAPTOR M3G and FERRET (CAC No. MF9141). (ADAMS Accession No. ML18204A010)

6. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014.

7. NRC email to Westinghouse Electric Company, dated April 20, 2022, U.S. NRC Final Safety Evaluation for Westinghouse Topical Report WCAP-18124-NP-A, Revision 0, Supplement 1 P and WCAP-18124-NP-A, Revision 0, Supplement 1-NP, "Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials" (EPID: L-2020-TOP-0063). (ADAMS Accession No. ML22153A139)

8. NRC letter to NSPM, dated September 3, 2020, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Reactor Vessel Material Surveillance Capsule Withdrawal Schedules (EPID L-2020-LLL-0016). (ADAMS Accession No. ML20230A051)

9. NSPM Letter L-PI-22-005, Prairie Island Unit 1 Reactor Vessel Material Surveillance Program Report, dated March 7, 2022. (ADAMS Accession No. ML22067A148)

ENCLOSURE 1 ATTACHMENT 1 Proposed Technical Specification Changes (Mark-Up) 2 pages follow

Reporting Requirements 5.6 Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-7 Unit 2 - Amendment No. 226 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, OPPS arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, RCS Pressure and Temperature (P/T) Limits; LCO 3.4.6, RCS Loops - MODE 4; LCO 3.4.7, RCS Loops - MODE 5, Loops Filled; LCO 3.4.10, Pressurizer Safety Valves; LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature; LCO 3.4.13, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature; and LCO 3.5.3, ECCS - Shutdown.

b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves (includes any exemption granted by NRC to ASME Code Case N-514).

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

Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

Insert A

Insert A The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

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

2.

WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," July 2018, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, "Fluence Determination with Raptor-M3G and FERRET - Supplement for Extended Beltline Materials," May 2022, shall be used as an alternative to Section 2.2 of WCAP-14040-NP-A.

ENCLOSURE 1 ATTACHMENT 2 Revised Technical Specification Pages 3 pages follow

Reporting Requirements 5.6 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 5.6-7 Unit 2 - Amendment No. TBD 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, OPPS arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, RCS Pressure and Temperature (P/T) Limits; LCO 3.4.6, RCS Loops - MODE 4; LCO 3.4.7, RCS Loops - MODE 5, Loops Filled; LCO 3.4.10, Pressurizer Safety Valves; LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature; LCO 3.4.13, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature; and LCO 3.5.3, ECCS - Shutdown.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

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

2. WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with Raptor-M3G and FERRET - Supplement for Extended Beltline Materials, May 2022, shall be used as an alternative to Section 2.2 of WCAP-14040-NP-A.

Reporting Requirements 5.6 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 5.6-8 Unit 2 - Amendment No. TBD 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

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

Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG;

b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available),

and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

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; and

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

Reporting Requirements 5.6 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 5.6-9 Unit 2 - Amendment No. TBD 5.6 Reporting Requirements (continued) 5.6.8 Steam Generator Tube Inspection Report (continued)

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; and

f.

The results of any SG secondary side inspections.

5.6.8 EM Report When a report is required by Condition C or I of LCO 3.3.3, "Event Monitoring (EM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

ENCLOSURE 2 WCAP-18746-NP, Revision 1 Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation 124 pages follow

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 WCAP-18746-NP October 2022 Revision 1 Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

Reactor Vessel/Containment Vessel Design and Analysis Riley I. Benson*

Nuclear Operations October 2022 Reviewers:

D. Brett Lynch*

Approved:

Lynn A. Patterson*, Manager RV/CV Design and Analysis RV/CV Design and Analysis Arturo Miralles Ferrete*

Jesse J Klingensmith*, Manager Radiation Engineering and Analysis Radiation Engineering and Analysis

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ii WCAP-18746-NP October 2022 Revision 1 RECORD OF REVISION Revision Description Completed 0

Original Issue July 2022 1

NSPM-RV000-CN-ME-000001 was updated to correct words Axial and Tangential which were swapped in Tables 4-2, 5-2, F-1, and F-2 as documented in CAP IR-2022-9008.

October 2022

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii WCAP-18746-NP October 2022 Revision 1 TABLE OF CONTENTS LIST OF TABLES........................................................................................................................................ v LIST OF FIGURES................................................................................................................................... viii EXECUTIVE

SUMMARY

.......................................................................................................................... ix 1

INTRODUCTION........................................................................................................... 1-1 2

RADIATION ANALYSIS AND NEUTRON DOSIMETRY.......................................... 2-1

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS........................................................................... 2-2 2.3 NEUTRON DOSIMETRY.............................................................................................. 2-4 2.4 CALCULATIONAL UNCERTAINTIES........................................................................ 2-5 3

FRACTURE TOUGHNESS PROPERTIES.................................................................... 3-1 4

SURVEILLANCE DATA................................................................................................ 4-1 5

CHEMISTRY FACTORS................................................................................................ 5-1 6

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS... 6-1 6.1 OVERALL APPROACH................................................................................................. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT............................................................................................................ 6-1 6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS........................................... 6-5 6.4 BOLTUP TEMPERATURE REQUIREMENTS............................................................. 6-5 7

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE............................. 7-1 8

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.......... 8-1 9

HEATUP AMD COOLDOWN LIMITS APPLICABILITY AND MARGIN ASSESSMENT................................................................................................................ 9-1 10 REFERENCES.............................................................................................................. 10-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt)...................................................... A-1 APPENDIX B OTHER RCPB FERRITIC COMPONENTS................................................................. B-1 APPENDIX C UPPER-SHELF ENERGY EVALUATION................................................................... C-1 APPENDIX D PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION................................................................................................. D-1 APPENDIX E VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS................................................................................ E-1

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iv WCAP-18746-NP October 2022 Revision 1 APPENDIX F CREDIBILITY EVALUATION OF THE PRAIRIE ISLAND UNIT 2 SURVEILLANCE PROGRAM...................................................................................................................... F-1

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 v

WCAP-18746-NP October 2022 Revision 1 LIST OF TABLES Table 2-1 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 1............................................................................................. 2-8 Table 2-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 1......................................................................................... 2-10 Table 2-3 Calculated Surveillance Capsule Lead Factors for Unit 1............................................. 2-12 Table 2-4 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1............................................................................. 2-13 Table 2-5 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 1............................................................................. 2-14 Table 2-6 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1.............................................................................................. 2-15 Table 2-7 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 1........................................................................................................ 2-16 Table 2-8 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 1.................................................. 2-17 Table 2-9 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Material for Unit 1........................................................................... 2-18 Table 2-10 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 2........................................................................................... 2-19 Table 2-11 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 2......................................................................................... 2-21 Table 2-12 Calculated Surveillance Capsule Lead Factors for Unit 2............................................. 2-23 Table 2-13 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2............................................................................. 2-24 Table 2-14 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 2............................................................................. 2-25 Table 2-15 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2.............................................................................................. 2-26 Table 2-16 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 2........................................................................................................ 2-27 Table 2-17 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 2.................................................. 2-28 Table 2-18 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Materials for Unit 2.......................................................................... 2-29

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vi WCAP-18746-NP October 2022 Revision 1 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 1 Reactor Vessel Materials................................................................ 3-2 Table 3-2 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 2 Reactor Vessel Materials................................................................ 3-3 Table 3-3 Summary of Prairie Island Units 1 and 2 Reactor Vessel Closure Head and Vessel Flange Initial RTNDT Values......................................................................................................... 3-4 Table 4-1 Prairie Island Unit 1 Surveillance Capsule Data.............................................................. 4-2 Table 4-2 Prairie Island Unit 2 Surveillance Capsule Data.............................................................. 4-3 Table 5-1 Calculation of Prairie Island Unit 1 Chemistry Factors Using Surveillance Capsule Data

......................................................................................................................................... 5-2 Table 5-2 Calculation of Prairie Island Unit 2 Chemistry Factors Using Surveillance Capsule Data

......................................................................................................................................... 5-3 Table 5-3 Summary of Prairie Island Unit 1 Positions 1.1 and 2.1 Chemistry Factors................... 5-4 Table 5-4 Summary of Prairie Island Unit 2 Positions 1.1 and 2.1 Chemistry Factors................... 5-5 Table 7-1 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 1 Reactor Vessel at 54 EFPY................. 7-3 Table 7-2 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 2 Reactor Vessel at 54 EFPY................. 7-3 Table 7-3 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location.............................................. 7-4 Table 7-4 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location.............................................. 7-5 Table 7-5 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location.............................................. 7-6 Table 7-6 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location.............................................. 7-7 Table 8-1 Summary of the Limiting ART Values Used in the Generation of the Prairie Island Units 1 and 2 Heatup and Cooldown Curves at 54 EFPY............................................... 8-2 Table 8-2 Prairie Island Units 1 and 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)................................................................................ 8-6 Table 8-3 Prairie Island Units 1 and 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)......................................................................... 8-8 Table A-1 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Heatup Curves........... A-2 Table A-2 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Cooldown Curves..... A-3

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vii WCAP-18746-NP October 2022 Revision 1 Table C-1 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 1 Beltline Materials................................................................................................. C-3 Table C-2 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 2 Beltline Materials................................................................................................. C-4 Table D-1 RTPTS Calculations for the Prairie Island Unit 1 Reactor Vessel Materials at 54 EFPY. D-2 Table D-2 RTPTS Calculations for the Prairie Island Unit 2 Reactor Vessel Materials at 54 EFPY. D-3 Table D-3 Evaluation of Prairie Island Unit 1 ERG Limit Category............................................... D-4 Table D-4 Evaluation of Prairie Island Unit 2 ERG Limit Category............................................... D-5 Table E-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors.................................... E-9 Table E-2 Measured Sensor Activities and Reaction Rates for Surveillance Capsule V.............. E-10 Table E-3 Measured Sensor Activities and Reaction Rates for Surveillance Capsule T............... E-11 Table E-4 Measured Sensor Activities and Reaction Rates for Surveillance Capsule R............... E-12 Table E-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule P............... E-13 Table E-6 Least-Squares Evaluation of Dosimetry in Surveillance Capsule V (13° Position, Core Midplane, Irradiated During Cycle 1)........................................................................... E-14 Table E-7 Least-Squares Evaluation of Dosimetry in Surveillance Capsule T (23° Position, Core Midplane, Irradiated During Cycles 1 through 4)......................................................... E-15 Table E-8 Least-Squares Evaluation of Dosimetry in Surveillance Capsule R (13° Position, Core Midplane, Irradiated During Cycles 1 through 9)......................................................... E-16 Table E-9 Least-Squares Evaluation of Dosimetry in Surveillance Capsule P (23° Position, Core Midplane, Irradiated During Cycles 1 through 16)....................................................... E-17 Table E-10 Measured-to-Calculated (M/C) Reaction Rates - In-Vessel Capsules.......................... E-18 Table E-11 Best-Estimate-to-Calculated (BE/C) Exposure Rates - In-Vessel Capsules................. E-18 Table F-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Prairie Island Unit 2 Surveillance Data...................................................................................... F-3 Table F-2 Best-Fit Evaluation for Prairie Island Unit 2 Surveillance Materials............................. F-4 Table F-3 Calculation of Residual vs. Fast Fluence for Prairie Island Unit 2................................. F-5

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 viii WCAP-18746-NP October 2022 Revision 1 LIST OF FIGURES Figure 2-1 Arrangement of Surveillance Capsules in the Prairie Island Units 1 and 2 Reactor Vessels

....................................................................................................................................... 2-30 Figure 2-2 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Core Midplane

....................................................................................................................................... 2-31 Figure 2-3 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Nozzle Centerline

....................................................................................................................................... 2-32 Figure 2-4 Prairie Island Units 1 and 2 Section View of the Reactor Geometry at =33°............. 2-33 Figure 8-1 Prairie Island Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100oF/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/KIc)....................................................................................................... 8-4 Figure 8-2 Prairie Island Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc).......................................................................... 8-5 Figure 9-1 Heatup Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws........................................... 9-2 Figure 9-2 Cooldown Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws........................................... 9-3 Figure C-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 1.............................................. C-5 Figure C-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 2.............................................. C-6

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ix WCAP-18746-NP October 2022 Revision 1 EXECUTIVE

SUMMARY

This report presents the evaluation of the Prairie Island Units 1 and 2 reactor pressure vessels (RPV) with respect to reactor vessel integrity, particularly with consideration of the Prairie Island Units 1 and 2 Capsule N testing results and RAPTOR-M3G fluence analysis as documented in Section 2. Prairie Island Units 1 and 2 have been approved for license extension for a total of 60 years of operation; thus, the evaluations in this report are projected through 54 effective full-power years (EFPY), which is deemed end-of-license extension (EOLE). Note that Capsule N from Prairie Island Unit 2 has not been analyzed at the time of issuance of this report, but a revision is planned to incorporate those results.

A summary of results for the Prairie Island Units 1 and 2 reactor vessel integrity evaluations are provided below. Based on the results presented herein, it is concluded that the Prairie Island Units 1 and 2 RPV will continue to meet RPV integrity regulatory requirements through the extended period of operation.

Although, reactor vessel integrity requirements continue to be met, the incorporation of the Prairie Island Units 1 and 2 Capsule N testing results and RAPTOR-M3G fluence analysis resulted in an increased limiting Adjusted Reference Temperature (ART) values for Prairie Island Units 1 and 2 through 54 EFPY.

In order to address this, heatup and cooldown Pressure-Temperature (P-T) limit curves were generated based on the revised ART values with additional margin and using the more current methodologies in WCAP-14040-A, Revision 4 and WCAP-18124-NP-A.

The P-T limit curves were generated for 54 EFPY using the KIc methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix G, which is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4. Heatup rates of 60 and 100F/hr, and cooldown rates of 0 (steady-state), 20, 40, 60, and 100F/hr were used to generate the P-T limit curves, with the flange requirements and without margins for instrumentation errors. The Prairie Island Units 1 and 2 End of License Extension (EOLE) corresponding to 60 years of operation is 54 EFPY. The EOLE P-T limit curves without margins for instrumentation errors can be found in Figure 8-1 and Figure 8-2. The P-T limit curves currently contained in the Pressure-Temperature Limit Report (PTLR) were generated with the more restrictive methodology contained in Revision 2 of WCAP-14040-A. Therefore, the curves generated in this report in Figure 8-1 and Figure 8-2 are compared to the PTLR P-T limits curves to determine if they remain applicable. The comparisons of the curves are shown in Figure 9-1 and Figure 9-2 which determined the current PTLR curves are bounding through 54 EFPY.

Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 54 EFPY.

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

Appendix C contains an upper-shelf energy (USE) evaluation for all Prairie Island Units 1 and 2 reactor vessel beltline and extended beltline materials. Per Appendix C, all beltline and extended beltline materials are projected to maintain USE values above the 50 ft-lb screening criterion per 10 CFR 50 Appendix G at 54 EFPY.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 x

WCAP-18746-NP October 2022 Revision 1 Appendix D contains a pressurized thermal shock (PTS) evaluation for all Prairie Island Units 1 and 2 reactor vessel beltline and extended beltline materials. Per Appendix D, all beltline and extended beltline materials have projected RTPTS values below the screening criteria set forth in 10 CFR 50.61. Additionally, Prairie Island Units 1 and 2 will remain in Category I of the Emergency Response Guidelines through 54 EFPY.

Appendix E contains the validation of the radiation transport models based on neutron dosimetry measurements for Unit 2. The validation of the radiation transport models based on neutron dosimetry measurements for Unit 1 are contained in WCAP-18660-NP [17].

Appendix F contains the credibility evaluation of the Prairie Island Unit 2 surveillance program.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-1 WCAP-18746-NP October 2022 Revision 1 1

INTRODUCTION The purpose of this report is to evaluate the Prairie Island Units 1 and 2 beltline materials with consideration of the testing results of the reactor vessel surveillance program Capsule N from Unit 1 to determine the impact on the Pressure-Temperature Limit Report (PTLR) P-T limit curves. Note that Capsule N from Prairie Island Unit 2 is scheduled for analysis in the near future and a revision to this report is planned to incorporate those results. The Units 1 and 2 beltline materials are also evaluated to determine their reference temperature for pressurized thermal shock (RTPTS) and upper-shelf energy (USE) values at end of license extension (EOLE), which corresponds to 54 effective full-power years (EFPY).

Reference nil-ductility transition temperature (RTNDT) increases, and the USE decreases as the material is exposed to fast-neutron irradiation. To find the most limiting RTNDT and USE at any time period in the reactor's life, Regulatory Guide 1.99 (RG 1.99), Revision 2 [1] is used to calculate the RTNDT and USE percent decease due to the associated radiation exposure. The resulting RTNDT values are used to adjust the unirradiated RTNDT (RTNDT(U)) in order to satisfy the requirements of 10 CFR Part 50.61 [10], the PTS Rule, and to verify/calculate P-T limit curves in accordance with the requirements of 10 CFR Part 50, Appendix G [4]. (Note, the methodology to calculate RTPTS is stipulated in 10 CFR Part 50.61; however, it is identical to RG 1.99.) The resulting limiting USE values are used to satisfy the requirements of 10 CFR 50, Appendix G.

The incorporation of the Unit 1 Capsule N surveillance data resulted in an increased limiting Adjusted Reference Temperature (ART). Therefore, new Prairie Island Units 1 and 2 heatup and cooldown P-T limit curves for 54 EFPY were developed. The heatup and cooldown P-T limit curves documented in this report were generated using the most limiting ART values (plus additional uncertainties to account for future perturbations such as the Unit 2 Surveillance Capsule N results) and the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [2]. Specifically, the KIc stress intensity factors and the less restrictive Circ-Flaw methodology of the 1998 through the 2000 Addenda Edition of ASME Code,Section XI, Appendix G [3] were used, when applicable. The KIc curve is a lower bound static fracture toughness curve based on crack initiation toughness. The limiting material is indexed to the KIc curve so that allowable stress intensity factors can be obtained for the material as a function of temperature.

Allowable operating limits are then determined using the allowable stress intensity factors. The P-T limit curves were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [4] have been incorporated in the P-T limit curves.

The Prairie Island Units 1 and 2 PTLR currently implements P-T limit curves developed in WCAP-14780

[9]. (Note, the curves in WCAP-14780 were originally developed for Unit 1 through 35 EFPY but have since been shown to bound Unit 2 as well and have been extended to 54 EFPY.) WCAP-14780, consistent with Revision 2 of WCAP-14040-NP-A used the more restrictive KIa stress intensity factors and Axial-Flaw methodology. KIa is a toughness based on the lower bound of crack arrest toughness, and the Axial-Flaw methodology was used even though the limiting ART values are located in a circumferential weld.

Because of the different methodologies, the new curves developed herein are compared to those currently in the PTLR to determine which are bounding.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-1 WCAP-18746-NP October 2022 Revision 1 2

RADIATION ANALYSIS AND NEUTRON DOSIMETRY

2.1 INTRODUCTION

Two discrete ordinates (Sn) transport analyses were performed for the Prairie Island Units 1 and 2 reactors to determine the neutron radiation environment within the reactor pressure vessels and surveillance capsules. In these analyses, neutron exposure parameters in terms of fast neutron (E > 1.0 MeV) fluence and iron atom displacements (dpa) were established on a plant-and fuel-cycle-specific basis. The dosimetry analysis documented in WCAP-18660-NP [17], Unit 1, and Appendix E, Unit 2, show that the +/-20% (1) acceptance criteria specified in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [14], is met, based on comparison with the five in-vessel surveillance capsules tested to-date for Prairie Island Unit 1 and the four in-vessel surveillance capsules tested to-date for Prairie Island Unit 2, respectively. As noted in Section 1, the Prairie Island Unit 2 Capsule N removed after the completion of Unit 2 Cycle 31 has not been analyzed at the time of this revision. When the evaluation of Prairie Island Unit 2 Capsule N is available, the updated dosimetry analysis results will be used to update this document accordingly. Comparisons of the results from the dosimetry evaluations with the analytical predictions served to validate the plant-specific neutron transport calculations. These validated calculations subsequently form the basis for projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY).

The use of fast neutron (E > 1.0 MeV) fluence to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. However, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy-dependent damage function for data correlation, ASTM E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, [12] recommends reporting displacements per iron atom along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy-dependent dpa function to be used for this evaluation is specified in ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom [13]. The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has been promulgated in Revision 2 to Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials [1].

All of the calculations and dosimetry evaluations described in this section were based on nuclear cross-section data derived from ENDF/B-VI. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [14]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET [15].

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-2 WCAP-18746-NP October 2022 Revision 1 2.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Prairie Island Units 1 and 2 reactor vessels is shown in Figure 2-1. Six irradiation capsules located between the thermal shield and vessel wall are included in each reactor design that constitutes the reactor vessel surveillance program. Capsules S, T, V, N, P, and R are located at azimuthal angles of 57°, 67°, 77°, 237°, 247°, and 257°, respectively. These full-core positions correspond to the following octant symmetric locations represented in Figure 2-2: 13° from the core cardinal axes (for the 77° and 257°), 23° from the core cardinal axes (for the 67° and 247°), and 33° from the core cardinal axes (for the 57° and 237°). The stainless-steel specimen containers are approximately 1-inch square in cross-section and are approximately 63 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core.

From a neutronic standpoint, the surveillance capsules and associated support structures are significant.

The presence of these materials has a significant effect on both the spatial distribution of neutron exposure rate and the neutron spectrum in the vicinity of the capsules. However, the capsules are far enough apart that they do not interfere with one another. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the Prairie Island Units 1 and 2 reactor vessels and surveillance capsules, plant-specific 3D forward transport calculations were carried out to directly solve for the space-and energy-dependent neutron exposure rate, (r,,z,E).

For the Prairie Island Units 1 and 2 transport calculations, the model depicted in Figure 2-2 through Figure 2-4 was utilized. The model contained a representation of the reactor core, the reactor internals, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing these analytical models, nominal design dimensions were generally employed for the various structural components. In addition, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a plant-and fuel-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.

A section view of the model is shown in Figure 2-4. Figure 2-4 shows the RAPTOR-M3G quadrant model at the 33° azimuth displaying the 33° surveillance capsule. Both models extended radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation more than fourteen (14) feet below the active fuel to about fourteen (14) feet above the active fuel.

The model consists of 186 radial mesh, 200 azimuthal mesh, and 435 axial mesh. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the RAPTOR-M3G calculations was set at a value of 0.001.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-3 WCAP-18746-NP October 2022 Revision 1 The core power distributions used in the plant-specific transport analyses for the first 32 fuel cycles at each of Prairie Island Units 1 and 2 included plant-and cycle-dependent fuel assembly initial enrichments, burnups, radial and axial power distributions. Actual operating characteristics through Cycle 32 have been evaluated for each unit; projections beyond Cycle 32 for each unit were based on the plant-specific Cycle 32 spatial power distributions with a 10% bias on the peripheral and re-entrant corners, water temperatures, and reactor power level as directed by Xcel Energy. The plant-and cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions were used to develop spatial-and energy-dependent core source distributions averaged over each individual fuel cycle for each unit. Therefore, the results from the neutron transport calculations provided data in terms of unit-fuel-cycle-averaged neutron exposure rate, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting these analyses were carried out using the RAPTOR-M3G discrete ordinates code and the BUGLE-96 cross-section library [5]. The BUGLE-96 library provides a coupled 47-neutron, 20-gamma-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 angular discretization was modeled with an S16 order of angular quadrature.

Results of the discrete ordinates transport analyses pertinent to the surveillance capsule evaluations are provided in Table 2-1 through Table 2-3 and Table 2-10 through Table 2-12 for Units 1 and 2, respectively.

In Table 2-1 and Table 2-10, the calculated fast neutron fluence rate and fluence (E > 1.0 MeV) are provided at the geometric center of the capsules, as a function of irradiation time for the Prairie Island Unit 1 and Unit 2 reactors, respectively. Similar data presented in terms of iron atom displacement rate and integrated iron atom displacements are given in Table 2-2 and Table 2-11 for Units 1 and 2, respectively.

In Table 2-3 and Table 2-12, lead factors associated with surveillance capsules are provided as a function of operating time for the Prairie Island Units 1 and 2 reactors, respectively. The lead factor is defined as the ratio of the neutron fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the maximum neutron fluence (E > 1.0 MeV) at the pressure vessel clad/base metal interface.

Neutron exposure data pertinent to the Unit 1 pressure vessel clad/base metal interface are given in Table 2-4 and Table 2-5 for neutron fluence (E > 1.0 MeV) rate and fluence (E > 1.0 MeV), respectively, and in Table 2-6 and Table 2-7 for dpa/s and dpa, respectively. Neutron exposure data pertinent to the Unit 2 pressure vessel clad/base metal interface are given in Table 2-13 and Table 2-14 for neutron fluence (E > 1.0 MeV) rate and fluence (E > 1.0 MeV), respectively, and in Table 2-15 and Table 2-16 for dpa/s and dpa, respectively. In each case, the data are provided for each operating cycle of the Prairie Island Units 1 and 2 reactors, respectively. Neutron fluence (E > 1.0 MeV) and dpa are also projected to future operating times extending to 60 EFPY for each unit. The vessel exposure data are presented in terms of the maximum exposure experienced by the pressure vessel at azimuthal angles of 0°, 15°, 30°, and 45°, and at the azimuthal location providing the maximum exposure relative to the core cardinal axes.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-4 WCAP-18746-NP October 2022 Revision 1 In Table 2-8 and Table 2-9, maximum projected fluences and dpa, respectively, of the various pressure vessel materials of Unit 1 are given. In Table 2-17 and Table 2-18, maximum projected fluences and dpa, respectively, of the various pressure vessel materials of Unit 2 are given.

These data tabulations include both plant-and fuel-cycle-specific calculated neutron exposures at the end of Cycle 32 and projections to 60 EFPY. The projections beyond Cycle 32 for both units were based on the plant-specific Cycle 32 spatial power distributions with a 10% bias on the peripheral and re-entrant corners, water temperatures, and reactor power level.

2.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures reported in Section 2.2 is demonstrated by a direct comparison against the measured sensor reaction rates and a least-squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serve to validate the calculated results, only the average of the direct comparison of measured-to-calculated results for the five previously analyzed surveillance capsules removed from Prairie Island Unit 1, Capsules V, P, R, S, and N, is provided in this section of the report. Since the evaluation of the Unit 2 Capsule N has not been completed yet, the validation of the transport model for Unit 2 using the neutron dosimetry of the previously withdrawn capsules is described in Appendix E. This report will be updated accordingly when the evaluation of the Unit 2 Capsule N is completed.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsules V, P, R, S, and N [17] that were withdrawn from the Unit 1 reactor, is summarized below.

Reaction M/C Reaction Rate Average

% Std. Dev.

63Cu (n,) 60Co 0.95 7.7 54Fe (n,p) 54Mn 0.86 6.7 58Ni (n,p) 58Co 0.91 5.6 238U(Cd) (n,f) 137Cs 0.97 7.6 237Np(Cd) (n,f) 137Cs 1.06 9.8 Average of M/C Results 0.94 9.8 The average measured-to-calculated (M/C) reaction rate ratios for the threshold reactions of the five previously withdrawn surveillance capsules for Prairie Island Unit 1 range from 0.86 to 1.06, and the average M/C ratio is 0.94 9.8% (1). This direct comparison falls within the 20% criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 2.2; therefore, the calculations for those cycles are deemed applicable for Prairie Island Unit 1.

As shown in Appendix E, the average measured-to-calculated (M/C) reaction rate ratio for the threshold reactions of the four previously withdrawn surveillance capsules from Prairie Island Unit 2 is 0.99 9.0%

(1). This direct comparison falls within the 20% criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 2.2; therefore, the calculations for those cycles are deemed applicable for Prairie Island Unit 2.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-5 WCAP-18746-NP October 2022 Revision 1 2.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Prairie Island Units 1 and 2 surveillance capsules and reactor pressure vessels is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

1. Simulator Benchmark Comparisons: Comparisons of calculations with measurements from simulator benchmarks, including the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL) and the VENUS-1 Experiment.

2. Operating Reactor and Calculational Benchmarks: Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment. Also considered are comparisons of calculations to results published in the NRC fluence calculation benchmark.

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

4. Plant-Specific Benchmarking: Comparisons of the plant-specific calculations with all available dosimetry results from the Prairie Island Units 1 and 2 surveillance programs.

The first phase of the methods qualification (simulator benchmark comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (operating reactor and calculational benchmark comparisons) addressed uncertainties in these additional areas that are primarily methods-related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations, as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Prairie Island Units 1 and 2 analyses was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Prairie Island Units 1 and 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. This comparison is used only as a check and is not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures described in Section 2.2. This comparison will be completed once again for Unit 2 when the evaluation of Unit 2 Capsule N is completed. This report will be updated accordingly when that is completed.

The following table summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Westinghouse Report WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET [15].

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-6 WCAP-18746-NP October 2022 Revision 1 Description Capsule and Vessel IR Simulator Benchmark Comparisons 3%

Operating Reactor and Calculational Benchmarks 5%

Analytic Uncertainty Analysis 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5%

Net Calculational Uncertainty 13%

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 NRC-issued Safety Evaluation for WCAP-18124-NP-A appears in Section A of [15]. The NRC identified two Limitations and Conditions associated with the application of RAPTOR-M3G and FERRET, which are reproduced here for convenience:

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 the 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 the 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 discrepancy should be disqualified.

The neutron exposure values applicable to the surveillance capsules and the maximum reactor pressure vessel neutron exposure values used to derive the surveillance capsule lead factors are completely covered by the benchmarking and uncertainty analyses in WCAP-18124-NP-A. Note, however, that this report does contain neutron exposure values for materials that are outside the qualification basis of WCAP-18124-NP-A (i.e. extended beltline materials). For the materials considered to be located in the extended beltline region, a comprehensive analytical uncertainty analysis applicable to the Prairie Island Units 1 and 2 RPV extended beltline region is summarized in WCAP18124-NP-A, Revision 0, Supplement 1-P/NP [19]. Note that the NRC has issued the Final Safety Evaluation Report for WCAP18124-NP-A, Revision 0, Supplement 1-P/NP. All RPV extended beltline calculations for Prairie Island Units 1 and 2 were performed using the WCAP-18124-NP-A Revision 0 Supplement 1-P/NP methodology.

Limitation #2 applies in situations where the least-squares analysis is used to adjust the calculated values of neutron exposure. In this report, the least-squares analysis is provided only as a supplemental check on

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-7 WCAP-18746-NP October 2022 Revision 1 the results of the dosimetry evaluation. The least-squares analysis was not used to modify the calculated surveillance capsule or reactor pressure vessel neutron exposure. Therefore, Limitation #2 does not apply.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-8 WCAP-18746-NP October 2022 Revision 1 Table 2-1 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence Rate (n/cm2-s) 13° 23° 33° 1

1.40 1.40 1.38E+11 7.87E+10 7.59E+10 2

0.78 2.18 1.43E+11 8.53E+10 8.25E+10 3

0.86 3.03 1.53E+11 8.47E+10 7.90E+10 4

0.89 3.92 1.50E+11 8.91E+10 8.60E+10 5

0.99 4.91 1.55E+11 8.72E+10 8.29E+10 6

0.87 5.79 1.57E+11 8.96E+10 8.57E+10 7

1.01 6.80 1.38E+11 8.47E+10 9.03E+10 8

0.89 7.69 1.70E+11 9.61E+10 9.30E+10 9

0.94 8.63 1.34E+11 8.91E+10 8.85E+10 10 0.92 9.55 1.78E+11 9.07E+10 8.15E+10 11 0.93 10.48 1.83E+11 1.06E+11 9.80E+10 12 1.18 11.65 1.32E+11 9.06E+10 8.77E+10 13 1.24 12.89 9.65E+10 7.01E+10 6.70E+10 14 1.21 14.11 7.82E+10 5.70E+10 5.66E+10 15 1.25 15.36 7.91E+10 5.59E+10 5.54E+10 16 1.29 16.65 9.15E+10 6.86E+10 6.29E+10 17 1.47 18.12 9.34E+10 6.88E+10 6.13E+10 18 1.55 19.68 8.21E+10 5.66E+10 5.35E+10 19 1.21 20.89 8.25E+10 6.71E+10 6.41E+10 20 1.61 22.50 8.97E+10 6.58E+10 5.97E+10 21 1.60 24.09 8.88E+10 5.82E+10 5.52E+10 22 1.72 25.81 9.07E+10 6.26E+10 6.19E+10 23 1.36 27.17 9.97E+10 6.72E+10 6.37E+10 24 1.61 28.78 8.79E+10 5.96E+10 5.60E+10 25 1.43 30.21 8.92E+10 6.34E+10 6.25E+10 26 1.42 31.63 9.10E+10 6.08E+10 5.84E+10 27 1.33 32.96 8.95E+10 5.59E+10 5.36E+10 28 1.74 34.71 8.79E+10 5.60E+10 5.37E+10 29 1.69 36.40 9.11E+10 6.01E+10 5.87E+10 30 1.79 38.19 8.87E+10 6.13E+10 6.30E+10 31 1.87 40.06 9.16E+10 5.93E+10 5.81E+10 32 2.03 42.08 8.77E+10 5.88E+10 5.96E+10

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-9 WCAP-18746-NP October 2022 Revision 1 Table 2-1 (continued) Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence (n/cm2)

V (13°)

P (23°)

R (13°)

S (33°)

N (33°)

T (23°)

1 1.40 1.40 6.09E+18 3.47E+18 6.09E+18 3.35E+18 3.35E+18 3.47E+18 2

0.78 2.18 5.57E+18 9.60E+18 5.38E+18 5.38E+18 5.57E+18 3

0.86 3.03 7.86E+18 1.37E+19 7.51E+18 7.51E+18 7.86E+18 4

0.89 3.92 1.04E+19 1.79E+19 9.93E+18 9.93E+18 1.04E+19 5

0.99 4.91 1.31E+19 2.28E+19 1.25E+19 1.25E+19 1.31E+19 6

0.87 5.79 2.71E+19 1.49E+19 1.49E+19 1.55E+19 7

1.01 6.80 3.15E+19 1.78E+19 1.78E+19 1.83E+19 8

0.89 7.69 3.63E+19 2.04E+19 2.04E+19 2.09E+19 9

0.94 8.63 4.02E+19 2.30E+19 2.30E+19 2.36E+19 10 0.92 9.55 2.54E+19 2.54E+19 2.62E+19 11 0.93 10.48 2.82E+19 2.82E+19 2.93E+19 12 1.18 11.65 3.15E+19 3.15E+19 3.27E+19 13 1.24 12.89 3.41E+19 3.41E+19 3.54E+19 14 1.21 14.11 3.63E+19 3.63E+19 3.76E+19 15 1.25 15.36 3.85E+19 3.85E+19 3.98E+19 16 1.29 16.65 4.10E+19 4.10E+19 4.26E+19 17 1.47 18.12 4.39E+19 4.39E+19 4.58E+19 18 1.55 19.68 4.65E+19 4.86E+19 19 1.21 20.89 4.90E+19 5.12E+19 20 1.61 22.50 5.20E+19 5.45E+19 21 1.60 24.09 5.48E+19 5.74E+19 22 1.72 25.81 5.81E+19 6.08E+19 23 1.36 27.17 6.08E+19 6.37E+19 24 1.61 28.78 6.37E+19 6.67E+19 25 1.43 30.21 6.65E+19 6.96E+19 26 1.42 31.63 6.91E+19 7.23E+19 27 1.33 32.96 7.14E+19 7.47E+19 28 1.74 34.71 7.44E+19 7.78E+19 29 1.69 36.40 7.75E+19 8.10E+19 30 1.79 38.19 8.10E+19 8.44E+19 31 1.87 40.06 8.45E+19 8.79E+19 32 2.03 42.08 9.17E+19 48 1.04E+20 51 1.10E+20 54 1.16E+20 60 1.28E+20

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-10 WCAP-18746-NP October 2022 Revision 1 Table 2-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Iron Atom Displacement Rate (dpa/s) 13° 23° 33° 1

1.40 1.40 2.53E-10 1.38E-10 1.34E-10 2

0.78 2.18 2.61E-10 1.50E-10 1.46E-10 3

0.86 3.03 2.80E-10 1.49E-10 1.39E-10 4

0.89 3.92 2.75E-10 1.56E-10 1.52E-10 5

0.99 4.91 2.84E-10 1.53E-10 1.46E-10 6

0.87 5.79 2.87E-10 1.57E-10 1.51E-10 7

1.01 6.80 2.53E-10 1.49E-10 1.59E-10 8

0.89 7.69 3.12E-10 1.69E-10 1.64E-10 9

0.94 8.63 2.44E-10 1.56E-10 1.56E-10 10 0.92 9.55 3.27E-10 1.59E-10 1.43E-10 11 0.93 10.48 3.37E-10 1.86E-10 1.73E-10 12 1.18 11.65 2.40E-10 1.58E-10 1.55E-10 13 1.24 12.89 1.75E-10 1.22E-10 1.18E-10 14 1.21 14.11 1.42E-10 9.90E-11 9.93E-11 15 1.25 15.36 1.43E-10 9.72E-11 9.71E-11 16 1.29 16.65 1.66E-10 1.19E-10 1.11E-10 17 1.47 18.12 1.69E-10 1.20E-10 1.08E-10 18 1.55 19.68 1.49E-10 9.84E-11 9.37E-11 19 1.21 20.89 1.49E-10 1.17E-10 1.13E-10 20 1.61 22.50 1.63E-10 1.14E-10 1.05E-10 21 1.60 24.09 1.61E-10 1.01E-10 9.68E-11 22 1.72 25.81 1.65E-10 1.09E-10 1.09E-10 23 1.36 27.17 1.81E-10 1.17E-10 1.12E-10 24 1.61 28.78 1.59E-10 1.04E-10 9.82E-11 25 1.43 30.21 1.62E-10 1.10E-10 1.10E-10 26 1.42 31.63 1.65E-10 1.06E-10 1.03E-10 27 1.33 32.96 1.63E-10 9.74E-11 9.40E-11 28 1.74 34.71 1.60E-10 9.76E-11 9.42E-11 29 1.69 36.40 1.66E-10 1.05E-10 1.03E-10 30 1.79 38.19 1.61E-10 1.07E-10 1.11E-10 31 1.87 40.06 1.66E-10 1.03E-10 1.02E-10 32 2.03 42.08 1.59E-10 1.02E-10 1.05E-10

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-11 WCAP-18746-NP October 2022 Revision 1 Table 2-2 (continued) Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Displacements (dpa)

V (13°)

P (23°)

R (13°)

S (33°)

N (33°)

T (23°)

1 1.40 1.40 1.12E-02 6.09E-03 1.12E-02 5.91E-03 5.91E-03 6.09E-03 2

0.78 2.18 9.76E-03 1.76E-02 9.49E-03 9.49E-03 9.76E-03 3

0.86 3.03 1.38E-02 2.52E-02 1.32E-02 1.32E-02 1.38E-02 4

0.89 3.92 1.82E-02 3.29E-02 1.75E-02 1.75E-02 1.82E-02 5

0.99 4.91 2.29E-02 4.18E-02 2.21E-02 2.21E-02 2.29E-02 6

0.87 5.79 4.97E-02 2.62E-02 2.62E-02 2.73E-02 7

1.01 6.80 5.78E-02 3.13E-02 3.13E-02 3.20E-02 8

0.89 7.69 6.65E-02 3.59E-02 3.59E-02 3.67E-02 9

0.94 8.63 7.38E-02 4.06E-02 4.06E-02 4.14E-02 10 0.92 9.55 4.47E-02 4.47E-02 4.60E-02 11 0.93 10.48 4.98E-02 4.98E-02 5.14E-02 12 1.18 11.65 5.55E-02 5.55E-02 5.73E-02 13 1.24 12.89 6.02E-02 6.02E-02 6.21E-02 14 1.21 14.11 6.40E-02 6.40E-02 6.59E-02 15 1.25 15.36 6.78E-02 6.78E-02 6.97E-02 16 1.29 16.65 7.23E-02 7.23E-02 7.46E-02 17 1.47 18.12 7.73E-02 7.73E-02 8.01E-02 18 1.55 19.68 8.19E-02 8.49E-02 19 1.21 20.89 8.62E-02 8.94E-02 20 1.61 22.50 9.15E-02 9.52E-02 21 1.60 24.09 9.64E-02 1.00E-01 22 1.72 25.81 1.02E-01 1.06E-01 23 1.36 27.17 1.07E-01 1.11E-01 24 1.61 28.78 1.12E-01 1.17E-01 25 1.43 30.21 1.17E-01 1.21E-01 26 1.42 31.63 1.22E-01 1.26E-01 27 1.33 32.96 1.26E-01 1.30E-01 28 1.74 34.71 1.31E-01 1.36E-01 29 1.69 36.40 1.36E-01 1.41E-01 30 1.79 38.19 1.43E-01 1.47E-01 31 1.87 40.06 1.49E-01 1.53E-01 32 2.03 42.08 1.60E-01 48 1.81E-01 51 1.91E-01 54 2.02E-01 60 2.23E-01

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-12 WCAP-18746-NP October 2022 Revision 1 Table 2-3 Calculated Surveillance Capsule Lead Factors for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Lead Factor 13° 23° 33° 1

1.40 1.40 2.98(a) 1.70 1.64 2

0.78 2.18 3.04 1.76 1.70 3

0.86 3.03 3.06 1.75 1.67 4

0.89 3.92 3.07 1.77 1.70 5

0.99 4.91 3.07 1.76(b) 1.69 6

0.87 5.79 3.07 1.76 1.69 7

1.01 6.80 3.11 1.80 1.75 8

0.89 7.69 3.10 1.79 1.74 9

0.94 8.63 3.08(c) 1.80 1.76 10 0.92 9.55 3.07 1.77 1.72 11 0.93 10.48 3.07 1.77 1.71 12 1.18 11.65 3.06 1.80 1.73 13 1.24 12.89 3.07 1.83 1.76 14 1.21 14.11 3.08 1.86 1.79 15 1.25 15.36 3.09 1.88 1.81 16 1.29 16.65 3.10 1.91 1.84 17 1.47 18.12 3.12 1.94 1.86(d) 18 1.55 19.68 3.13 1.96 1.87 19 1.21 20.89 3.14 1.99 1.90 20 1.61 22.50 3.15 2.01 1.92 21 1.60 24.09 3.14 2.01 1.92 22 1.72 25.81 3.15 2.02 1.93 23 1.36 27.17 3.15 2.02 1.93 24 1.61 28.78 3.15 2.03 1.94 25 1.43 30.21 3.15 2.04 1.95 26 1.42 31.63 3.16 2.05 1.95 27 1.33 32.96 3.16 2.04 1.95 28 1.74 34.71 3.16 2.04 1.95 29 1.69 36.40 3.15 2.04 1.95 30 1.79 38.19 3.16 2.05 1.97 31 1.87 40.06 3.15 2.05 1.97(e) 32 2.03 42.08 3.15 2.05 1.97 48(f) 3.15 2.05 1.99 51 3.15 2.06 2.00 54 3.15 2.06 2.00 60 3.15 2.06 2.01 Notes:

(a) Capsule V was removed after Cycle 1.

(b) Capsule P was removed after Cycle 5.

(c) Capsule R was removed after Cycle 9.

(d) Capsule S was removed after Cycle 17.

(e) Capsule N was removed after Cycle 31.

(f) The projections beyond Cycle 32 are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-13 WCAP-18746-NP October 2022 Revision 1 Table 2-4 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence Rate (n/cm2-s) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.40 1.40 4.61E+10 2.74E+10 1.82E+10 1.60E+10 4.64E+10

-73 2

0.78 2.18 4.77E+10 2.95E+10 2.03E+10 1.73E+10 4.79E+10 121 3

0.86 3.03 4.96E+10 2.93E+10 1.88E+10 1.70E+10 4.98E+10

-1 4

0.89 3.92 4.81E+10 2.90E+10 2.03E+10 1.75E+10 4.83E+10 3

5 0.99 4.91 5.05E+10 3.01E+10 1.97E+10 1.78E+10 5.08E+10 67 6

0.87 5.79 5.06E+10 3.06E+10 2.03E+10 1.75E+10 5.09E+10

-73 7

1.01 6.80 4.13E+10 2.68E+10 2.07E+10 1.92E+10 4.16E+10 5

8 0.89 7.69 5.57E+10 3.24E+10 2.20E+10 1.81E+10 5.60E+10 5

9 0.94 8.63 4.60E+10 2.66E+10 2.09E+10 1.73E+10 4.62E+10 3

10 0.92 9.55 5.84E+10 3.36E+10 1.95E+10 1.86E+10 5.87E+10 5

11 0.93 10.48 5.98E+10 3.52E+10 2.36E+10 1.73E+10 6.02E+10 5

12 1.18 11.65 4.43E+10 2.65E+10 2.10E+10 1.63E+10 4.45E+10 5

13 1.24 12.89 2.94E+10 2.00E+10 1.61E+10 1.44E+10 2.96E+10 5

14 1.21 14.11 2.39E+10 1.64E+10 1.35E+10 1.27E+10 2.40E+10 73 15 1.25 15.36 2.36E+10 1.63E+10 1.31E+10 1.26E+10 2.37E+10 5

16 1.29 16.65 2.76E+10 1.91E+10 1.54E+10 1.29E+10 2.77E+10 3

17 1.47 18.12 2.75E+10 1.95E+10 1.51E+10 1.23E+10 2.76E+10 5

18 1.55 19.68 2.51E+10 1.69E+10 1.29E+10 1.17E+10 2.52E+10 5

19 1.21 20.89 2.31E+10 1.77E+10 1.55E+10 1.36E+10 2.32E+10 67 20 1.61 22.50 2.75E+10 1.92E+10 1.50E+10 1.25E+10 2.76E+10 71 21 1.60 24.09 2.94E+10 1.80E+10 1.33E+10 1.19E+10 2.96E+10 5

22 1.72 25.81 2.87E+10 1.88E+10 1.49E+10 1.34E+10 2.88E+10 67 23 1.36 27.17 3.17E+10 2.04E+10 1.54E+10 1.39E+10 3.18E+10 67 24 1.61 28.78 2.76E+10 1.81E+10 1.36E+10 1.21E+10 2.78E+10 67 25 1.43 30.21 2.73E+10 1.82E+10 1.48E+10 1.30E+10 2.74E+10 5

26 1.42 31.63 2.83E+10 1.87E+10 1.40E+10 1.24E+10 2.85E+10

-73 27 1.33 32.96 2.86E+10 1.82E+10 1.29E+10 1.17E+10 2.88E+10

-73 28 1.74 34.71 2.79E+10 1.79E+10 1.29E+10 1.18E+10 2.80E+10

-73 29 1.69 36.40 2.93E+10 1.86E+10 1.40E+10 1.22E+10 2.95E+10

-73 30 1.79 38.19 2.73E+10 1.82E+10 1.47E+10 1.37E+10 2.75E+10

-73 31 1.87 40.06 2.94E+10 1.85E+10 1.38E+10 1.22E+10 2.96E+10

-73 32 2.03 42.08 2.80E+10 1.79E+10 1.40E+10 1.30E+10 2.81E+10

-73

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-14 WCAP-18746-NP October 2022 Revision 1 Table 2-5 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence (n/cm2) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.4 1.4 2.04E+18 1.21E+18 8.04E+17 7.08E+17 2.05E+18

-73 2

0.78 2.18 3.14E+18 1.89E+18 1.28E+18 1.11E+18 3.16E+18

-73 3

0.86 3.03 4.47E+18 2.66E+18 1.78E+18 1.57E+18 4.49E+18

-3 4

0.89 3.92 5.81E+18 3.48E+18 2.35E+18 2.06E+18 5.84E+18

-3 5

0.99 4.91 7.38E+18 4.38E+18 2.96E+18 2.61E+18 7.42E+18

-1 6

0.87 5.79 8.77E+18 5.21E+18 3.52E+18 3.09E+18 8.81E+18

-1 7

1.01 6.8 1.01E+19 6.07E+18 4.18E+18 3.71E+18 1.01E+19

-1 8

0.89 7.69 1.16E+19 6.97E+18 4.80E+18 4.22E+18 1.17E+19 1

9 0.94 8.63 1.30E+19 7.76E+18 5.42E+18 4.73E+18 1.31E+19 1

10 0.92 9.55 1.47E+19 8.74E+18 5.98E+18 5.27E+18 1.48E+19 3

11 0.93 10.48 1.65E+19 9.77E+18 6.68E+18 5.78E+18 1.65E+19 3

12 1.18 11.65 1.81E+19 1.08E+19 7.45E+18 6.38E+18 1.82E+19 3

13 1.24 12.89 1.93E+19 1.15E+19 8.09E+18 6.95E+18 1.94E+19 3

14 1.21 14.11 2.02E+19 1.22E+19 8.60E+18 7.43E+18 2.03E+19 3

15 1.25 15.36 2.11E+19 1.28E+19 9.12E+18 7.93E+18 2.12E+19 3

16 1.29 16.65 2.22E+19 1.36E+19 9.75E+18 8.45E+18 2.23E+19 3

17 1.47 18.12 2.35E+19 1.45E+19 1.04E+19 9.02E+18 2.36E+19 3

18 1.55 19.68 2.47E+19 1.53E+19 1.11E+19 9.59E+18 2.48E+19 3

19 1.21 20.89 2.56E+19 1.60E+19 1.17E+19 1.01E+19 2.57E+19 3

20 1.61 22.5 2.70E+19 1.69E+19 1.24E+19 1.07E+19 2.71E+19 5

21 1.6 24.09 2.84E+19 1.78E+19 1.31E+19 1.13E+19 2.86E+19 5

22 1.72 25.81 3.00E+19 1.88E+19 1.39E+19 1.21E+19 3.01E+19 5

23 1.36 27.17 3.13E+19 1.97E+19 1.45E+19 1.26E+19 3.15E+19 5

24 1.61 28.78 3.27E+19 2.06E+19 1.52E+19 1.33E+19 3.29E+19 5

25 1.43 30.21 3.39E+19 2.14E+19 1.59E+19 1.38E+19 3.41E+19 5

26 1.42 31.63 3.52E+19 2.23E+19 1.65E+19 1.44E+19 3.54E+19 5

27 1.33 32.96 3.64E+19 2.30E+19 1.70E+19 1.49E+19 3.66E+19 5

28 1.74 34.71 3.79E+19 2.40E+19 1.78E+19 1.55E+19 3.81E+19 5

29 1.69 36.4 3.95E+19 2.49E+19 1.85E+19 1.62E+19 3.97E+19 5

30 1.79 38.19 4.10E+19 2.59E+19 1.93E+19 1.69E+19 4.12E+19 5

31 1.87 40.06 4.27E+19 2.70E+19 2.01E+19 1.77E+19 4.29E+19 3

32 2.03 42.08 4.45E+19 2.81E+19 2.10E+19 1.85E+19 4.47E+19 3

48(a) 5.02E+19 3.17E+19 2.39E+19 2.11E+19 5.05E+19 3

51 5.32E+19 3.36E+19 2.53E+19 2.24E+19 5.34E+19 3

54 5.61E+19 3.54E+19 2.67E+19 2.38E+19 5.63E+19 3

60 6.19E+19 3.90E+19 2.96E+19 2.65E+19 6.22E+19 3

Notes:

(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-15 WCAP-18746-NP October 2022 Revision 1 Table 2-6 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Displacement Rate (dpa/s) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.40 1.40 7.56E-11 4.56E-11 3.00E-11 2.61E-11 7.59E-11

-73 2

0.78 2.18 7.83E-11 4.88E-11 3.30E-11 2.81E-11 7.87E-11 119 3

0.86 3.03 8.14E-11 4.97E-11 3.11E-11 2.77E-11 8.17E-11

-1 4

0.89 3.92 7.89E-11 4.93E-11 3.36E-11 2.85E-11 7.92E-11 3

5 0.99 4.91 8.30E-11 5.07E-11 3.25E-11 2.89E-11 8.33E-11 67 6

0.87 5.79 8.30E-11 5.10E-11 3.35E-11 2.85E-11 8.34E-11

-73 7

1.01 6.80 6.79E-11 4.55E-11 3.42E-11 3.13E-11 6.82E-11 5

8 0.89 7.69 9.14E-11 5.51E-11 3.63E-11 2.95E-11 9.18E-11 5

9 0.94 8.63 7.54E-11 4.51E-11 3.44E-11 2.82E-11 7.57E-11 3

10 0.92 9.55 9.59E-11 5.71E-11 3.23E-11 3.02E-11 9.63E-11 5

11 0.93 10.48 9.83E-11 5.98E-11 3.90E-11 2.83E-11 9.87E-11 5

12 1.18 11.65 7.26E-11 4.48E-11 3.46E-11 2.67E-11 7.29E-11 5

13 1.24 12.89 4.82E-11 3.37E-11 2.65E-11 2.35E-11 4.84E-11 5

14 1.21 14.11 3.92E-11 2.74E-11 2.22E-11 2.06E-11 3.94E-11 73 15 1.25 15.36 3.87E-11 2.75E-11 2.17E-11 2.04E-11 3.88E-11 5

16 1.29 16.65 4.52E-11 3.22E-11 2.53E-11 2.09E-11 4.54E-11 3

17 1.47 18.12 4.50E-11 3.28E-11 2.49E-11 2.00E-11 4.52E-11 5

18 1.55 19.68 4.11E-11 2.84E-11 2.12E-11 1.90E-11 4.12E-11 5

19 1.21 20.89 3.79E-11 2.98E-11 2.56E-11 2.22E-11 3.81E-11 67 20 1.61 22.50 4.51E-11 3.22E-11 2.46E-11 2.04E-11 4.53E-11 73 21 1.60 24.09 4.81E-11 3.04E-11 2.19E-11 1.93E-11 4.84E-11 5

22 1.72 25.81 4.70E-11 3.16E-11 2.45E-11 2.18E-11 4.72E-11 67 23 1.36 27.17 5.18E-11 3.43E-11 2.53E-11 2.26E-11 5.20E-11 67 24 1.61 28.78 4.52E-11 3.04E-11 2.24E-11 1.97E-11 4.54E-11 67 25 1.43 30.21 4.47E-11 3.06E-11 2.44E-11 2.11E-11 4.49E-11 5

26 1.42 31.63 4.64E-11 3.10E-11 2.30E-11 2.01E-11 4.66E-11

-73 27 1.33 32.96 4.68E-11 3.01E-11 2.11E-11 1.90E-11 4.70E-11

-73 28 1.74 34.71 4.56E-11 2.97E-11 2.12E-11 1.91E-11 4.57E-11

-73 29 1.69 36.40 4.79E-11 3.08E-11 2.30E-11 1.99E-11 4.82E-11

-73 30 1.79 38.19 4.47E-11 3.02E-11 2.42E-11 2.22E-11 4.49E-11

-3 31 1.87 40.06 4.81E-11 3.08E-11 2.26E-11 1.98E-11 4.84E-11

-73 32 2.03 42.08 4.58E-11 2.97E-11 2.29E-11 2.10E-11 4.60E-11 61

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-16 WCAP-18746-NP October 2022 Revision 1 Table 2-7 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Displacements (dpa) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.40 1.40 3.34E-03 2.01E-03 1.32E-03 1.15E-03 3.35E-03

-73 2

0.78 2.18 5.15E-03 3.14E-03 2.10E-03 1.81E-03 5.17E-03

-73 3

0.86 3.03 7.33E-03 4.49E-03 2.94E-03 2.55E-03 7.36E-03

-3 4

0.89 3.92 9.54E-03 5.87E-03 3.88E-03 3.35E-03 9.58E-03

-3 5

0.99 4.91 1.21E-02 7.44E-03 4.89E-03 4.25E-03 1.22E-02

-1 6

0.87 5.79 1.44E-02 8.85E-03 5.81E-03 5.04E-03 1.45E-02

-1 7

1.01 6.80 1.66E-02 1.03E-02 6.90E-03 6.04E-03 1.66E-02

-1 8

0.89 7.69 1.91E-02 1.18E-02 7.92E-03 6.87E-03 1.92E-02

-1 9

0.94 8.63 2.14E-02 1.32E-02 8.94E-03 7.70E-03 2.14E-02

-1 10 0.92 9.55 2.41E-02 1.48E-02 9.88E-03 8.58E-03 2.42E-02 3

11 0.93 10.48 2.70E-02 1.66E-02 1.10E-02 9.41E-03 2.71E-02 3

12 1.18 11.65 2.97E-02 1.83E-02 1.23E-02 1.04E-02 2.98E-02 3

13 1.24 12.89 3.16E-02 1.96E-02 1.33E-02 1.13E-02 3.17E-02 3

14 1.21 14.11 3.31E-02 2.06E-02 1.42E-02 1.21E-02 3.32E-02 3

15 1.25 15.36 3.46E-02 2.17E-02 1.51E-02 1.29E-02 3.48E-02 3

16 1.29 16.65 3.65E-02 2.30E-02 1.61E-02 1.38E-02 3.66E-02 3

17 1.47 18.12 3.85E-02 2.45E-02 1.72E-02 1.47E-02 3.87E-02 5

18 1.55 19.68 4.06E-02 2.59E-02 1.83E-02 1.56E-02 4.07E-02 5

19 1.21 20.89 4.20E-02 2.71E-02 1.93E-02 1.65E-02 4.22E-02 5

20 1.61 22.50 4.42E-02 2.87E-02 2.05E-02 1.75E-02 4.44E-02 5

21 1.60 24.09 4.66E-02 3.02E-02 2.16E-02 1.85E-02 4.68E-02 5

22 1.72 25.81 4.92E-02 3.19E-02 2.29E-02 1.96E-02 4.94E-02 5

23 1.36 27.17 5.13E-02 3.33E-02 2.40E-02 2.06E-02 5.16E-02 5

24 1.61 28.78 5.36E-02 3.49E-02 2.51E-02 2.16E-02 5.38E-02 5

25 1.43 30.21 5.56E-02 3.63E-02 2.62E-02 2.25E-02 5.59E-02 5

26 1.42 31.63 5.77E-02 3.76E-02 2.72E-02 2.34E-02 5.79E-02 5

27 1.33 32.96 5.96E-02 3.89E-02 2.81E-02 2.42E-02 5.99E-02 5

28 1.74 34.71 6.21E-02 4.05E-02 2.93E-02 2.53E-02 6.24E-02 5

29 1.69 36.40 6.47E-02 4.22E-02 3.05E-02 2.63E-02 6.50E-02 5

30 1.79 38.19 6.72E-02 4.39E-02 3.18E-02 2.76E-02 6.75E-02 5

31 1.87 40.06 7.00E-02 4.57E-02 3.32E-02 2.87E-02 7.03E-02 5

32 2.03 42.08 7.30E-02 4.76E-02 3.46E-02 3.01E-02 7.33E-02 5

48(a) 8.23E-02 5.37E-02 3.93E-02 3.43E-02 8.27E-02 3

51 8.71E-02 5.67E-02 4.17E-02 3.65E-02 8.74E-02 3

54 9.18E-02 5.98E-02 4.41E-02 3.87E-02 9.22E-02 3

60 1.01E-01 6.60E-02 4.88E-02 4.30E-02 1.02E-01 3

Notes:

(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-17 WCAP-18746-NP October 2022 Revision 1 Table 2-8 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 1 Material Fast Fluence (n/cm2) 42.1 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 2.63E+19 3.00E+19 3.37E+19 3.74E+19 Intermediate Shell Forging 4.47E+19 5.05E+19 5.63E+19 6.22E+19 Lower Shell Forging 4.37E+19 4.95E+19 5.53E+19 6.12E+19 Inlet Nozzle to Nozzle Shell Weld - Lowest Extent(a) 2.56E+16 2.96E+16 3.36E+16 3.76E+16 Upper to Intermediate Shell Weld 2.84E+19 3.23E+19 3.63E+19 4.02E+19 Intermediate to Lower Shell Weld 4.38E+19 4.95E+19 5.53E+19 6.12E+19 Lower Shell to Lower Closure Head Weld 1.49E+16 1.70E+16 1.92E+16 2.14E+16 Notes:

(a) The outlet nozzle weld fluence is bounded by the inlet nozzle location fluence.

(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-18 WCAP-18746-NP October 2022 Revision 1 Table 2-9 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Material for Unit 1 Material Iron Atom Displacements (dpa) 42.1 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 4.33E-02 4.93E-02 5.54E-02 6.15E-02 Intermediate Shell Forging 7.33E-02 8.27E-02 9.22E-02 1.02E-01 Lower Shell Forging 7.16E-02 8.10E-02 9.05E-02 1.00E-01 Inlet Nozzle to Nozzle Shell Weld

- Lowest Extent(a) 1.25E-04 1.43E-04 1.62E-04 1.80E-04 Upper to Intermediate Shell Weld 4.67E-02 5.31E-02 5.96E-02 6.61E-02 Intermediate to Lower Shell Weld 7.17E-02 8.10E-02 9.06E-02 1.00E-01 Lower Shell to Lower Closure Head Weld 7.60E-05 8.70E-05 9.81E-05 1.09E-04 Notes:

(a) The outlet nozzle weld dpa is bounded by the inlet nozzle location dpa.

(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-19 WCAP-18746-NP October 2022 Revision 1 Table 2-10 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence Rate (n/cm2-s) 13° 23° 33° 1

1.39 1.39 1.36E+11 7.79E+10 7.48E+10 2

0.87 2.26 1.45E+11 8.68E+10 8.40E+10 3

0.89 3.15 1.50E+11 8.91E+10 8.71E+10 4

0.98 4.13 1.51E+11 8.84E+10 8.46E+10 5

0.92 5.05 1.58E+11 9.03E+10 8.74E+10 6

0.99 6.04 1.49E+11 7.93E+10 7.12E+10 7

1.01 7.05 1.43E+11 8.20E+10 7.62E+10 8

0.90 7.95 1.25E+11 7.61E+10 7.06E+10 9

0.85 8.80 1.82E+11 9.43E+10 8.63E+10 10 0.92 9.71 1.58E+11 9.42E+10 9.12E+10 11 1.09 10.80 1.48E+11 8.63E+10 8.25E+10 12 1.08 11.88 1.26E+11 8.51E+10 8.57E+10 13 1.26 13.15 9.40E+10 6.46E+10 6.25E+10 14 1.33 14.47 8.76E+10 6.11E+10 6.11E+10 15 1.38 15.86 8.99E+10 5.67E+10 5.52E+10 16 1.38 17.24 9.43E+10 7.13E+10 6.69E+10 17 1.55 18.78 8.79E+10 6.56E+10 5.87E+10 18 1.48 20.27 8.37E+10 5.78E+10 5.47E+10 19 1.30 21.57 8.10E+10 6.50E+10 6.10E+10 20 1.56 23.13 9.53E+10 6.84E+10 6.04E+10 21 1.52 24.64 8.87E+10 6.30E+10 5.54E+10 22 1.48 26.12 8.86E+10 6.17E+10 5.81E+10 23 1.37 27.49 8.95E+10 6.00E+10 5.77E+10 24 1.72 29.21 8.69E+10 5.77E+10 5.48E+10 25 1.44 30.66 8.60E+10 5.94E+10 5.99E+10 26 1.68 32.34 8.87E+10 5.82E+10 5.74E+10 27 1.24 33.58 9.32E+10 6.10E+10 6.06E+10 28 1.65 35.24 9.35E+10 6.04E+10 5.70E+10 29 1.64 36.88 9.36E+10 6.14E+10 6.10E+10 30 1.85 38.73 8.86E+10 6.10E+10 6.21E+10 31 1.91 40.64 9.20E+10 5.91E+10 5.77E+10 32 1.95 42.59 9.15E+10 6.04E+10 6.01E+10

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-20 WCAP-18746-NP October 2022 Revision 1 Table 2-10 (continued) Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence (n/cm2)

V (13°)

T (23°)

R (13°)

P (23°)

N (33°)

S (33°)

1 1.39 1.39 5.98E+18 3.42E+18 5.98E+18 3.42E+18 3.29E+18 3.29E+18 2

0.87 2.26 5.81E+18 9.97E+18 5.81E+18 5.59E+18 5.59E+18 3

0.89 3.15 8.30E+18 1.42E+19 8.30E+18 8.03E+18 8.03E+18 4

0.98 4.13 1.10E+19 1.88E+19 1.10E+19 1.06E+19 1.06E+19 5

0.92 5.05 2.34E+19 1.37E+19 1.32E+19 1.32E+19 6

0.99 6.04 2.81E+19 1.61E+19 1.54E+19 1.54E+19 7

1.01 7.05 3.26E+19 1.88E+19 1.78E+19 1.78E+19 8

0.90 7.95 3.62E+19 2.09E+19 1.98E+19 1.98E+19 9

0.85 8.80 4.11E+19 2.34E+19 2.22E+19 2.22E+19 10 0.92 9.71 2.62E+19 2.48E+19 2.48E+19 11 1.09 10.80 2.91E+19 2.76E+19 2.76E+19 12 1.08 11.88 3.20E+19 3.05E+19 3.05E+19 13 1.26 13.15 3.46E+19 3.30E+19 3.30E+19 14 1.33 14.47 3.72E+19 3.56E+19 3.56E+19 15 1.38 15.86 3.96E+19 3.80E+19 3.80E+19 16 1.38 17.24 4.27E+19 4.09E+19 4.09E+19 17 1.55 18.78 4.38E+19 4.38E+19 18 1.48 20.27 4.64E+19 4.64E+19 19 1.30 21.57 4.89E+19 4.89E+19 20 1.56 23.13 5.18E+19 5.18E+19 21 1.52 24.64 5.45E+19 5.45E+19 22 1.48 26.12 5.72E+19 5.72E+19 23 1.37 27.49 5.97E+19 5.97E+19 24 1.72 29.21 6.27E+19 6.27E+19 25 1.44 30.66 6.54E+19 6.54E+19 26 1.68 32.34 6.84E+19 6.84E+19 27 1.24 33.58 7.08E+19 7.08E+19 28 1.65 35.24 7.38E+19 7.38E+19 29 1.64 36.88 7.69E+19 7.69E+19 30 1.85 38.73 8.06E+19 8.06E+19 31 1.91 40.64 8.41E+19 8.41E+19 32 1.95 42.59 8.77E+19 48 9.90E+19 51 1.05E+20 54 1.11E+20 60 1.24E+20

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-21 WCAP-18746-NP October 2022 Revision 1 Table 2-11 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Iron Atom Displacement Rate (dpa/s) 13° 23° 33° 1

1.39 1.39 2.49E-10 1.37E-10 1.32E-10 2

0.87 2.26 2.66E-10 1.52E-10 1.48E-10 3

0.89 3.15 2.75E-10 1.56E-10 1.54E-10 4

0.98 4.13 2.77E-10 1.55E-10 1.49E-10 5

0.92 5.05 2.90E-10 1.58E-10 1.54E-10 6

0.99 6.04 2.73E-10 1.39E-10 1.25E-10 7

1.01 7.05 2.62E-10 1.44E-10 1.34E-10 8

0.90 7.95 2.29E-10 1.33E-10 1.24E-10 9

0.85 8.80 3.34E-10 1.66E-10 1.52E-10 10 0.92 9.71 2.89E-10 1.65E-10 1.61E-10 11 1.09 10.80 2.70E-10 1.51E-10 1.45E-10 12 1.08 11.88 2.30E-10 1.49E-10 1.51E-10 13 1.26 13.15 1.71E-10 1.13E-10 1.10E-10 14 1.33 14.47 1.59E-10 1.06E-10 1.07E-10 15 1.38 15.86 1.63E-10 9.88E-11 9.69E-11 16 1.38 17.24 1.71E-10 1.24E-10 1.17E-10 17 1.55 18.78 1.59E-10 1.14E-10 1.03E-10 18 1.48 20.27 1.52E-10 1.00E-10 9.59E-11 19 1.30 21.57 1.47E-10 1.13E-10 1.07E-10 20 1.56 23.13 1.73E-10 1.19E-10 1.06E-10 21 1.52 24.64 1.61E-10 1.09E-10 9.72E-11 22 1.48 26.12 1.61E-10 1.07E-10 1.02E-10 23 1.37 27.49 1.62E-10 1.04E-10 1.01E-10 24 1.72 29.21 1.58E-10 1.00E-10 9.61E-11 25 1.44 30.66 1.56E-10 1.03E-10 1.05E-10 26 1.68 32.34 1.61E-10 1.01E-10 1.01E-10 27 1.24 33.58 1.69E-10 1.06E-10 1.06E-10 28 1.65 35.24 1.70E-10 1.05E-10 9.99E-11 29 1.64 36.88 1.70E-10 1.07E-10 1.07E-10 30 1.85 38.73 1.61E-10 1.06E-10 1.09E-10 31 1.91 40.64 1.67E-10 1.03E-10 1.01E-10 32 1.95 42.59 1.66E-10 1.05E-10 1.05E-10

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-22 WCAP-18746-NP October 2022 Revision 1 Table 2-11 (continued) Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Displacements (dpa)

V (13°)

T (23°)

R (13°)

P (23°)

N (33°)

S (33°)

1 1.39 1.39 1.10E-02 6.00E-03 1.10E-02 6.00E-03 5.79E-03 5.79E-03 2

0.87 2.26 1.02E-02 1.83E-02 1.02E-02 9.86E-03 9.86E-03 3

0.89 3.15 1.46E-02 2.60E-02 1.46E-02 1.42E-02 1.42E-02 4

0.98 4.13 1.93E-02 3.45E-02 1.93E-02 1.88E-02 1.88E-02 5

0.92 5.05 4.29E-02 2.39E-02 2.32E-02 2.32E-02 6

0.99 6.04 5.15E-02 2.83E-02 2.72E-02 2.72E-02 7

1.01 7.05 5.98E-02 3.29E-02 3.14E-02 3.14E-02 8

0.90 7.95 6.63E-02 3.66E-02 3.50E-02 3.50E-02 9

0.85 8.80 7.53E-02 4.11E-02 3.90E-02 3.90E-02 10 0.92 9.71 4.58E-02 4.37E-02 4.37E-02 11 1.09 10.80 5.10E-02 4.87E-02 4.87E-02 12 1.08 11.88 5.61E-02 5.38E-02 5.38E-02 13 1.26 13.15 6.06E-02 5.82E-02 5.82E-02 14 1.33 14.47 6.50E-02 6.27E-02 6.27E-02 15 1.38 15.86 6.93E-02 6.69E-02 6.69E-02 16 1.38 17.24 7.47E-02 7.20E-02 7.20E-02 17 1.55 18.78 7.71E-02 7.71E-02 18 1.48 20.27 8.16E-02 8.16E-02 19 1.30 21.57 8.60E-02 8.60E-02 20 1.56 23.13 9.12E-02 9.12E-02 21 1.52 24.64 9.58E-02 9.58E-02 22 1.48 26.12 1.01E-01 1.01E-01 23 1.37 27.49 1.05E-01 1.05E-01 24 1.72 29.21 1.10E-01 1.10E-01 25 1.44 30.66 1.15E-01 1.15E-01 26 1.68 32.34 1.20E-01 1.20E-01 27 1.24 33.58 1.24E-01 1.24E-01 28 1.65 35.24 1.30E-01 1.30E-01 29 1.64 36.88 1.35E-01 1.35E-01 30 1.85 38.73 1.42E-01 1.42E-01 31 1.91 40.64 1.48E-01 1.48E-01 32 1.95 42.59 1.54E-01 48 1.74E-01 51 1.85E-01 54 1.96E-01 60 2.18E-01

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-23 WCAP-18746-NP October 2022 Revision 1 Table 2-12 Calculated Surveillance Capsule Lead Factors for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Lead Factor 13° 23° 33° 1

1.39 1.39 3.03(a) 1.73 1.66 2

0.87 2.26 3.06 1.78 1.72 3

0.89 3.15 3.08 1.80 1.74 4

0.98 4.13 3.08 1.80(b) 1.74 5

0.92 5.05 3.07 1.79 1.73 6

0.99 6.04 3.07 1.76 1.68 7

1.01 7.05 3.11 1.78 1.70 8

0.90 7.95 3.09 1.78 1.69 9

0.85 8.80 3.08(c) 1.76 1.66 10 0.92 9.71 3.08 1.77 1.68 11 1.09 10.80 3.07 1.76 1.67 12 1.08 11.88 3.08 1.80 1.71 13 1.26 13.15 3.09 1.82 1.74 14 1.33 14.47 3.09 1.84 1.76 15 1.38 15.86 3.09 1.85 1.77 16 1.38 17.24 3.11 1.88(d) 1.80 17 1.55 18.78 3.12 1.92 1.83 18 1.48 20.27 3.13 1.93 1.84 19 1.30 21.57 3.14 1.97 1.87 20 1.56 23.13 3.15 1.99 1.89 21 1.52 24.64 3.16 2.01 1.89 22 1.48 26.12 3.16 2.02 1.90 23 1.37 27.49 3.16 2.02 1.91 24 1.72 29.21 3.16 2.02 1.91 25 1.44 30.66 3.16 2.02 1.92 26 1.68 32.34 3.15 2.02 1.92 27 1.24 33.58 3.15 2.02 1.92 28 1.65 35.24 3.15 2.02 1.92 29 1.64 36.88 3.15 2.03 1.93 30 1.85 38.73 3.16 2.04 1.94 31 1.91 40.64 3.15 2.03 1.94(e) 32 1.95 42.59 3.15 2.04 1.95 48(f) 3.15 2.04 1.96 51 3.15 2.04 1.97 54 3.15 2.04 1.97 60 3.15 2.04 1.98 Notes:

(a) Capsule V was removed after Cycle 1.

(b) Capsule T was removed after Cycle 4.

(c) Capsule R was removed after Cycle 9.

(d) Capsule P was removed after Cycle 16.

(e) Capsule N was removed after Cycle 31.

(f) The projections beyond Cycle 32 are based on Cycle 32 with a 10% bias on peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-24 WCAP-18746-NP October 2022 Revision 1 Table 2-13 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence Rate (n/cm2-s) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.39 1.39 4.49E+10 2.69E+10 1.77E+10 1.55E+10 4.49E+10

-9 2

0.87 2.26 4.75E+10 2.93E+10 2.01E+10 1.73E+10 4.75E+10 73 3

0.89 3.15 4.83E+10 2.97E+10 2.05E+10 1.80E+10 4.83E+10

-1 4

0.98 4.13 4.90E+10 3.00E+10 2.01E+10 1.72E+10 4.90E+10 5

5 0.92 5.05 5.19E+10 3.11E+10 2.06E+10 1.84E+10 5.19E+10

-3 6

0.99 6.04 5.03E+10 2.98E+10 1.74E+10 1.62E+10 5.03E+10 59 7

1.01 7.05 4.24E+10 2.84E+10 1.82E+10 1.75E+10 4.24E+10

-3 8

0.90 7.95 4.28E+10 2.53E+10 1.70E+10 1.47E+10 4.28E+10

-3 9

0.85 8.80 5.98E+10 3.53E+10 2.06E+10 1.92E+10 5.98E+10

-3 10 0.92 9.71 5.07E+10 3.13E+10 2.15E+10 1.85E+10 5.07E+10 5

11 1.09 10.80 5.06E+10 2.96E+10 1.95E+10 1.89E+10 5.06E+10 5

12 1.08 11.88 3.87E+10 2.60E+10 2.01E+10 1.89E+10 3.87E+10 5

13 1.26 13.15 2.96E+10 1.97E+10 1.50E+10 1.33E+10 2.96E+10 5

14 1.33 14.47 2.79E+10 1.84E+10 1.45E+10 1.29E+10 2.79E+10 73 15 1.38 15.86 2.95E+10 1.85E+10 1.32E+10 1.26E+10 2.95E+10 5

16 1.38 17.24 2.78E+10 2.02E+10 1.62E+10 1.41E+10 2.78E+10 5

17 1.55 18.78 2.61E+10 1.89E+10 1.45E+10 1.21E+10 2.61E+10 5

18 1.48 20.27 2.56E+10 1.76E+10 1.32E+10 1.19E+10 2.56E+10 5

19 1.30 21.57 2.27E+10 1.76E+10 1.48E+10 1.26E+10 2.27E+10 5

20 1.56 23.13 2.87E+10 2.03E+10 1.50E+10 1.19E+10 2.87E+10 67 21 1.52 24.64 2.69E+10 1.89E+10 1.38E+10 1.16E+10 2.69E+10 67 22 1.48 26.12 2.75E+10 1.86E+10 1.41E+10 1.24E+10 2.76E+10 67 23 1.37 27.49 2.87E+10 1.86E+10 1.38E+10 1.26E+10 2.87E+10 67 24 1.72 29.21 2.86E+10 1.80E+10 1.31E+10 1.14E+10 2.86E+10 3

25 1.44 30.66 2.76E+10 1.79E+10 1.41E+10 1.28E+10 2.76E+10

-73 26 1.68 32.34 2.93E+10 1.83E+10 1.36E+10 1.28E+10 2.93E+10

-73 27 1.24 33.58 3.02E+10 1.92E+10 1.44E+10 1.32E+10 3.02E+10

-73 28 1.65 35.24 2.97E+10 1.91E+10 1.36E+10 1.22E+10 2.97E+10

-3 29 1.64 36.88 2.88E+10 1.91E+10 1.44E+10 1.31E+10 2.88E+10

-3 30 1.85 38.73 2.76E+10 1.83E+10 1.46E+10 1.34E+10 2.76E+10

-73 31 1.91 40.64 2.96E+10 1.88E+10 1.37E+10 1.21E+10 2.96E+10

-73 32 1.95 42.59 2.93E+10 1.88E+10 1.42E+10 1.29E+10 2.93E+10

-3

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-25 WCAP-18746-NP October 2022 Revision 1 Table 2-14 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Fluence (n/cm2) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.39 1.39 1.98E+18 1.18E+18 7.78E+17 6.81E+17 1.98E+18

-9 2

0.87 2.26 3.25E+18 1.97E+18 1.32E+18 1.15E+18 3.25E+18

-3 3

0.89 3.15 4.60E+18 2.81E+18 1.90E+18 1.65E+18 4.60E+18

-3 4

0.98 4.13 6.12E+18 3.73E+18 2.52E+18 2.18E+18 6.12E+18

-1 5

0.92 5.05 7.62E+18 4.63E+18 3.11E+18 2.72E+18 7.62E+18

-3 6

0.99 6.04 9.15E+18 5.55E+18 3.65E+18 3.22E+18 9.15E+18

-1 7

1.01 7.05 1.05E+19 6.45E+18 4.23E+18 3.78E+18 1.05E+19

-1 8

0.9 7.95 1.17E+19 7.17E+18 4.71E+18 4.19E+18 1.17E+19

-1 9

0.85 8.8 1.33E+19 8.12E+18 5.26E+18 4.71E+18 1.33E+19

-1 10 0.92 9.71 1.48E+19 9.02E+18 5.89E+18 5.24E+18 1.48E+19

-1 11 1.09 10.8 1.65E+19 1.00E+19 6.56E+18 5.89E+18 1.65E+19

-1 12 1.08 11.88 1.78E+19 1.09E+19 7.24E+18 6.54E+18 1.78E+19 1

13 1.26 13.15 1.90E+19 1.17E+19 7.84E+18 7.07E+18 1.90E+19 3

14 1.33 14.47 2.02E+19 1.25E+19 8.45E+18 7.61E+18 2.02E+19 3

15 1.38 15.86 2.15E+19 1.33E+19 9.02E+18 8.16E+18 2.15E+19 3

16 1.38 17.24 2.27E+19 1.42E+19 9.73E+18 8.77E+18 2.27E+19 3

17 1.55 18.78 2.39E+19 1.51E+19 1.04E+19 9.36E+18 2.39E+19 3

18 1.48 20.27 2.51E+19 1.59E+19 1.11E+19 9.92E+18 2.51E+19 3

19 1.3 21.57 2.61E+19 1.66E+19 1.17E+19 1.04E+19 2.61E+19 3

20 1.56 23.13 2.75E+19 1.76E+19 1.24E+19 1.10E+19 2.75E+19 5

21 1.52 24.64 2.88E+19 1.85E+19 1.31E+19 1.16E+19 2.88E+19 5

22 1.48 26.12 3.00E+19 1.94E+19 1.37E+19 1.22E+19 3.00E+19 5

23 1.37 27.49 3.13E+19 2.02E+19 1.43E+19 1.27E+19 3.13E+19 5

24 1.72 29.21 3.28E+19 2.12E+19 1.50E+19 1.33E+19 3.28E+19 5

25 1.44 30.66 3.41E+19 2.20E+19 1.57E+19 1.39E+19 3.41E+19 5

26 1.68 32.34 3.56E+19 2.29E+19 1.64E+19 1.46E+19 3.56E+19 5

27 1.24 33.58 3.68E+19 2.37E+19 1.69E+19 1.51E+19 3.68E+19 5

28 1.65 35.24 3.83E+19 2.47E+19 1.77E+19 1.57E+19 3.83E+19 3

29 1.64 36.88 3.98E+19 2.57E+19 1.84E+19 1.64E+19 3.98E+19 3

30 1.85 38.73 4.15E+19 2.68E+19 1.92E+19 1.72E+19 4.15E+19 3

31 1.91 40.64 4.32E+19 2.79E+19 2.01E+19 1.79E+19 4.32E+19 3

32 1.95 42.59 4.50E+19 2.90E+19 2.09E+19 1.87E+19 4.50E+19 3

48(a) 5.05E+19 3.25E+19 2.36E+19 2.11E+19 5.05E+19 3

51 5.35E+19 3.45E+19 2.50E+19 2.24E+19 5.35E+19 3

54 5.66E+19 3.64E+19 2.65E+19 2.38E+19 5.66E+19 3

60 6.26E+19 4.03E+19 2.94E+19 2.64E+19 6.26E+19 3

Notes:

(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-26 WCAP-18746-NP October 2022 Revision 1 Table 2-15 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Displacement Rate (dpa/s) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.39 1.39 7.37E-11 4.44E-11 2.89E-11 2.52E-11 7.37E-11

-9 2

0.87 2.26 7.79E-11 4.84E-11 3.28E-11 2.82E-11 7.79E-11 73 3

0.89 3.15 7.92E-11 4.91E-11 3.35E-11 2.93E-11 7.92E-11

-1 4

0.98 4.13 8.04E-11 4.95E-11 3.27E-11 2.80E-11 8.04E-11 5

5 0.92 5.05 8.51E-11 5.14E-11 3.36E-11 3.00E-11 8.51E-11

-3 6

0.99 6.04 8.25E-11 4.93E-11 2.84E-11 2.64E-11 8.25E-11 57 7

1.01 7.05 6.95E-11 4.68E-11 2.97E-11 2.84E-11 6.95E-11

-3 8

0.90 7.95 7.01E-11 4.18E-11 2.77E-11 2.39E-11 7.01E-11

-3 9

0.85 8.80 9.82E-11 5.83E-11 3.37E-11 3.13E-11 9.82E-11

-3 10 0.92 9.71 8.32E-11 5.17E-11 3.51E-11 3.02E-11 8.32E-11 5

11 1.09 10.80 8.28E-11 4.88E-11 3.19E-11 3.07E-11 8.28E-11 5

12 1.08 11.88 6.34E-11 4.28E-11 3.28E-11 3.08E-11 6.34E-11 9

13 1.26 13.15 4.85E-11 3.23E-11 2.44E-11 2.16E-11 4.85E-11 5

14 1.33 14.47 4.56E-11 3.02E-11 2.36E-11 2.10E-11 4.56E-11 73 15 1.38 15.86 4.83E-11 3.05E-11 2.14E-11 2.05E-11 4.83E-11 9

16 1.38 17.24 4.55E-11 3.32E-11 2.64E-11 2.29E-11 4.55E-11 5

17 1.55 18.78 4.27E-11 3.10E-11 2.36E-11 1.97E-11 4.27E-11 5

18 1.48 20.27 4.18E-11 2.89E-11 2.15E-11 1.93E-11 4.18E-11 5

19 1.30 21.57 3.72E-11 2.89E-11 2.41E-11 2.05E-11 3.72E-11 5

20 1.56 23.13 4.70E-11 3.33E-11 2.44E-11 1.94E-11 4.70E-11 65 21 1.52 24.64 4.40E-11 3.10E-11 2.24E-11 1.89E-11 4.40E-11 65 22 1.48 26.12 4.50E-11 3.06E-11 2.29E-11 2.02E-11 4.52E-11 67 23 1.37 27.49 4.69E-11 3.05E-11 2.24E-11 2.05E-11 4.69E-11 67 24 1.72 29.21 4.67E-11 2.95E-11 2.13E-11 1.86E-11 4.67E-11 3

25 1.44 30.66 4.51E-11 2.94E-11 2.30E-11 2.08E-11 4.51E-11

-73 26 1.68 32.34 4.78E-11 3.01E-11 2.22E-11 2.07E-11 4.78E-11

-73 27 1.24 33.58 4.93E-11 3.16E-11 2.34E-11 2.15E-11 4.93E-11

-71 28 1.65 35.24 4.86E-11 3.14E-11 2.22E-11 1.98E-11 4.86E-11 61 29 1.64 36.88 4.71E-11 3.14E-11 2.34E-11 2.13E-11 4.71E-11

-3 30 1.85 38.73 4.51E-11 3.01E-11 2.37E-11 2.17E-11 4.51E-11 65 31 1.91 40.64 4.84E-11 3.09E-11 2.23E-11 1.97E-11 4.84E-11

-73 32 1.95 42.59 4.80E-11 3.09E-11 2.30E-11 2.09E-11 4.80E-11 63

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-27 WCAP-18746-NP October 2022 Revision 1 Table 2-16 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)

Total Time (EFPY)

Displacements (dpa) 0° 15° 30° 45° Maximum Elevation of Max.

(cm) 1 1.39 1.39 3.24E-03 1.95E-03 1.27E-03 1.11E-03 3.24E-03

-9 2

0.87 2.26 5.33E-03 3.26E-03 2.16E-03 1.87E-03 5.33E-03

-3 3

0.89 3.15 7.55E-03 4.63E-03 3.09E-03 2.69E-03 7.55E-03

-3 4

0.98 4.13 1.00E-02 6.16E-03 4.11E-03 3.56E-03 1.00E-02

-1 5

0.92 5.05 1.25E-02 7.65E-03 5.08E-03 4.43E-03 1.25E-02

-3 6

0.99 6.04 1.50E-02 9.16E-03 5.95E-03 5.24E-03 1.50E-02

-1 7

1.01 7.05 1.72E-02 1.07E-02 6.90E-03 6.15E-03 1.72E-02

-1 8

0.90 7.95 1.92E-02 1.18E-02 7.69E-03 6.83E-03 1.92E-02

-1 9

0.85 8.80 2.19E-02 1.34E-02 8.59E-03 7.67E-03 2.19E-02

-1 10 0.92 9.71 2.43E-02 1.49E-02 9.60E-03 8.54E-03 2.43E-02

-1 11 1.09 10.80 2.71E-02 1.66E-02 1.07E-02 9.59E-03 2.71E-02

-1 12 1.08 11.88 2.92E-02 1.80E-02 1.18E-02 1.06E-02 2.92E-02

-1 13 1.26 13.15 3.12E-02 1.93E-02 1.28E-02 1.15E-02 3.12E-02 3

14 1.33 14.47 3.31E-02 2.06E-02 1.38E-02 1.24E-02 3.31E-02 3

15 1.38 15.86 3.52E-02 2.19E-02 1.47E-02 1.33E-02 3.52E-02 3

16 1.38 17.24 3.72E-02 2.33E-02 1.59E-02 1.43E-02 3.72E-02 3

17 1.55 18.78 3.93E-02 2.49E-02 1.70E-02 1.52E-02 3.93E-02 3

18 1.48 20.27 4.12E-02 2.62E-02 1.80E-02 1.61E-02 4.12E-02 5

19 1.30 21.57 4.27E-02 2.74E-02 1.90E-02 1.70E-02 4.27E-02 5

20 1.56 23.13 4.50E-02 2.90E-02 2.02E-02 1.79E-02 4.50E-02 5

21 1.52 24.64 4.71E-02 3.05E-02 2.13E-02 1.88E-02 4.71E-02 5

22 1.48 26.12 4.92E-02 3.19E-02 2.24E-02 1.98E-02 4.92E-02 5

23 1.37 27.49 5.13E-02 3.33E-02 2.33E-02 2.07E-02 5.13E-02 5

24 1.72 29.21 5.38E-02 3.49E-02 2.45E-02 2.17E-02 5.38E-02 5

25 1.44 30.66 5.58E-02 3.62E-02 2.55E-02 2.26E-02 5.58E-02 5

26 1.68 32.34 5.84E-02 3.78E-02 2.67E-02 2.37E-02 5.84E-02 5

27 1.24 33.58 6.03E-02 3.90E-02 2.76E-02 2.45E-02 6.03E-02 5

28 1.65 35.24 6.28E-02 4.06E-02 2.88E-02 2.56E-02 6.28E-02 5

29 1.64 36.88 6.52E-02 4.23E-02 3.00E-02 2.67E-02 6.52E-02 5

30 1.85 38.73 6.79E-02 4.40E-02 3.13E-02 2.79E-02 6.79E-02 3

31 1.91 40.64 7.08E-02 4.59E-02 3.27E-02 2.91E-02 7.08E-02 3

32 1.95 42.59 7.37E-02 4.78E-02 3.41E-02 3.04E-02 7.37E-02 3

48(a) 8.27E-02 5.35E-02 3.84E-02 3.43E-02 8.27E-02 3

51 8.76E-02 5.67E-02 4.08E-02 3.65E-02 8.76E-02 3

54 9.26E-02 5.99E-02 4.32E-02 3.86E-02 9.26E-02 3

60 1.02E-01 6.63E-02 4.79E-02 4.30E-02 1.02E-01

-1 Notes:

(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-28 WCAP-18746-NP October 2022 Revision 1 Table 2-17 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 2 Material Fast Fluence (n/cm2) 42.6 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 2.63E+19 2.98E+19 3.37E+19 3.76E+19 Intermediate Shell Forging 4.50E+19 5.05E+19 5.66E+19 6.26E+19 Lower Shell Forging 4.43E+19 4.98E+19 5.58E+19 6.19E+19 Inlet Nozzle to Nozzle Shell Weld - Lowest Extent(a) 2.64E+16 3.01E+16 3.42E+16 3.82E+16 Upper to Intermediate Shell Weld 2.82E+19 3.20E+19 3.61E+19 4.03E+19 Intermediate to Lower Shell Weld 4.43E+19 4.98E+19 5.58E+19 6.19E+19 Lower Shell to Lower Closure Head Weld 1.52E+16 1.73E+16 1.95E+16 2.18E+16 Notes:

(a) The outlet nozzle weld fluence is bounded by the inlet nozzle location fluence.

(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-29 WCAP-18746-NP October 2022 Revision 1 Table 2-18 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Materials for Unit 2 Material Iron Atom Displacements (dpa) 42.6 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 4.32E-02 4.90E-02 5.54E-02 6.17E-02 Intermediate Shell Forging 7.37E-02 8.27E-02 9.26E-02 1.03E-01 Lower Shell Forging 7.25E-02 8.14E-02 9.13E-02 1.01E-01 Inlet Nozzle to Nozzle Shell Weld

- Lowest Extent(a) 1.26E-04 1.43E-04 1.62E-04 1.80E-04 Upper to Intermediate Shell Weld 4.65E-02 5.26E-02 5.94E-02 6.62E-02 Intermediate to Lower Shell Weld 7.25E-02 8.14E-02 9.13E-02 1.01E-01 Lower Shell to Lower Closure Head Weld 7.78E-05 8.82E-05 9.97E-05 1.11E-04 Notes:

(a) Outlet nozzle weld dpa is bounded by the inlet nozzle location dpa.

(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-30 WCAP-18746-NP October 2022 Revision 1 Figure 2-1 Arrangement of Surveillance Capsules in the Prairie Island Units 1 and 2 Reactor Vessels

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-31 WCAP-18746-NP October 2022 Revision 1 Figure 2-2 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Core Midplane

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-32 WCAP-18746-NP October 2022 Revision 1 Figure 2-3 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Nozzle Centerline

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-33 WCAP-18746-NP October 2022 Revision 1 Figure 2-4 Prairie Island Units 1 and 2 Section View of the Reactor Geometry at =33°

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-1 WCAP-18746-NP October 2022 Revision 1 3

FRACTURE TOUGHNESS PROPERTIES The requirements for reactor vessel integrity and P-T limit curve development are specified in 10 CFR 50, Appendix G [4]. 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 Prairie Island Unit 1 beltline materials traditionally included Nozzle (a.k.a. Upper) Shell Forging B, Intermediate Shell Forging C, Lower Shell Forging D, Nozzle Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2, and Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3. The Prairie Island Unit 2 beltline materials traditionally included Upper Shell Forging B, Intermediate Shell Forging C, Lower Shell Forging D, Upper Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2, and Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3. However, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [6], 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 to experience neutron fluence sufficient to cause embrittlement. The additional materials that exceed this fluence threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met. Since no additional materials are expected to reach this threshold by the EOLE per Table 2-8 (Unit 1) and Table 2-17 (Unit 2), no extended beltline materials are identified.

A summary of the best-estimate copper (Cu) and nickel (Ni), contents, in units of weight percent (wt. %),

as well as initial RTNDT and initial USE values for the reactor vessel beltline materials are provided in Table 3-1 and Table 3-2 for Prairie Island Unit 1 and 2, respectively. Table 3-3 contains a summary of the initial RTNDT values of the reactor vessel flange and reactor vessel closure head flange. These flange initial RTNDT values serve as input to the P-T limit curves flange-notch per Appendix G of 10 CFR 50 - See Section 6.3 for details.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-2 WCAP-18746-NP October 2022 Revision 1 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 1 Reactor Vessel Materials(a)

Reactor Vessel Material and Identification Number Heat Number Flux Type (Lot)

Chemical Composition Fracture Toughness Property Wt.

% Cu Wt.

Ni Initial RTNDT

(°F)

I

(°F)

Initial Upper-Shelf Energy (ft-lb)

Nozzle Shell Forging B 21744/38384 0.08 0.68

-4 0

84 Intermediate Shell Forging C 21918/38566 0.07 0.80 14 0

143 Lower Shell Forging D 21887/38530 0.07 0.66

-4 0

88 Nozzle Shell to Intermediate Shell Circumferential Weld Seam W2 2269 UM 89 (1180) 0.15 0.15 0(b) 17(b) 84 Intermediate Shell to Lower Shell Circumferential Weld Seam W3 1752 UM 89 (1230) 0.13 0.13

-13 0

78.5 Surveillance Materials Intermediate Shell Forging C 21918/38566 Surveillance Weld 1752 UM 89 (1230) 0.14 0.11 Notes:

(a) Values are measured values from the CMTRs or other Prairie Island Unit 1 fabrication records unless otherwise noted.

(b) The initial RTNDT of weld seam W2 is estimated based on Branch Technical Position 5-3 [7]; therefore, I = 17°F.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-3 WCAP-18746-NP October 2022 Revision 1 Table 3-2 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 2 Reactor Vessel Materials(a)

Reactor Vessel Material and Identification Number Heat Number Flux Type (Lot)

Chemical Composition Fracture Toughness Property Wt. %

Cu Wt. %

Ni Initial RTNDT

(°F)

I

(°F)

Initial Upper-Shelf Energy (ft-lb)

Upper Shell Forging B 22231/39088 0.07 0.73

-13 0

85 Intermediate Shell Forging C 22829 0.07 0.75 14 0

112 Lower Shell Forging D 22642 0.08 0.67

-4 0

108 Upper Shell to Intermediate Shell Circumferential Weld Seam W2 1752 UM 89 (1263) 0.13(b) 0.13(b)

-13(b) 0 78.5(b)

Intermediate Shell to Lower Shell Circumferential Weld Seam W3 2721 UM 89 (1263) 0.09 0.11

-31 0

103 Surveillance Materials Lower Shell Forging D 22642 Surveillance Weld 2721 UM 89 (1263) 0.08 0.08 Surveillance Weld Material from Prairie Island Unit 1 1752 UM 89 (1230) 0.14 0.11 Notes:

(a) Values are measured values from the CMTRs or other Prairie Island Unit 2 fabrication records unless otherwise noted.

(b) Consistent with Prairie Island Unit 1 Intermediate Shell to Lower Shell Circumferential Weld Seam W3.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-4 WCAP-18746-NP October 2022 Revision 1 Table 3-3 Summary of Prairie Island Units 1 and 2 Reactor Vessel Closure Head and Vessel Flange Initial RTNDT Values Reactor Vessel Material Unit 1 Initial RTNDT

(°F)

Unit 2 Initial RTNDT

(°F)

Replacement Reactor Vessel Closure Head

-70

-50 Vessel Flange

-4

-22

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-1 WCAP-18746-NP October 2022 Revision 1 4

SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2 [1], calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, surveillance data generated from surveillance programs at other plants which include Prairie Island Unit 1 or 2 reactor vessel material should also be considered when calculating Position 2.1 chemistry factors. Note that material from Heat # 1752 from the Prairie Island Unit 1 surveillance program in used in the Prairie Island Unit 2 reactor vessel.

The surveillance capsule forging material for Prairie Island Unit 1 is from Intermediate Shell Forging C.

Per WCAP-18660-NP [17], the surveillance forging data are deemed non-credible. The Prairie Island Unit 1 surveillance weld specimens were fabricated from weld material Heat # 1752 flux type UM 89, Lot # 1230.

This weld material is applicable to the Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3. Per WCAP-18660-NP, the surveillance weld data are deemed non-credible. Therefore, a reduced margin term cannot be utilized with the Position 2.1 chemistry factors in the ART or PTS calculations for these materials contained in Section 7 and Appendix D, respectively.

The surveillance capsule forging material for Prairie Island Unit 2 is from Lower Shell Forging D. The surveillance forging data are deemed non-credible in Appendix F; therefore, a reduced margin term cannot be utilized with the Position 2.1 chemistry factor in the Lower Shell Forging D ART or PTS calculations contained in Section 7 and Appendix D, respectively. The Prairie Island Unit 2 surveillance weld specimens were fabricated from weld material Heat # 2721 flux type UM 89, Lot # 1263. This weld material is applicable to the Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3.

The surveillance weld data are deemed credible in Appendix F; therefore, a reduced margin term will be utilized with the Position 2.1 chemistry factor in the ART and PTS calculations for these welds contained in Section 7 and Appendix D, respectively.

Table 4-1 and Table 4-2 summarize the surveillance data available for the Prairie Island Units 1 and 2 forging and weld materials that will be used in the calculation of the Position 2.1 chemistry factor values, respectively.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-2 WCAP-18746-NP October 2022 Revision 1 Table 4-1 Prairie Island Unit 1 Surveillance Capsule Data(a)

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

Measured 30 ft-lb Transition Temperature Shift

(°F)

Measured Upper-Shelf Energy Decrease

(%)

Intermediate Shell Forging C (Tangential)

V 0.609 31.9 2

P 1.31 23.2 10 R

4.02 96.0 9

S 4.39 101.6 10 N

8.45 165.5 13 Intermediate Shell Forging C (Axial)

V 0.609 48.8 0(b)

P 1.31 33.9 5

R 4.02 84.0 10 S

4.39 74.0 6

N 8.45 147.5 16 Surveillance Weld (Heat # 1752)

V 0.609 35.2 0(b)

P 1.31 45.6 0(b)

R 4.02 123.0 9

S 4.39 161.0 0(b)

N 8.45 219.0 4

Notes:

(a) Surveillance data from WCAP-18660-NP [17].

(b) An increase in USE was measured, which physically should not occur after irradiation. Therefore, a conservative 0% decrease value is used.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-3 WCAP-18746-NP October 2022 Revision 1 Table 4-2 Prairie Island Unit 2 Surveillance Capsule Data Material Capsule Capsule Fluence (x 1019 n/cm2, E > 1.0 MeV)

Measured 30 ft-lb Transition Temperature Shift

(°F)(a)

Measured Upper-Shelf Energy Decrease

(%)(a)

Lower Shell Forging D (Axial)

V 0.598 35.28 0

T 1.10 29.93 13.8 R

4.11 84.73 9.3 P

4.27 103.87 13 Lower Shell Forging D (Tangential)

V 0.598 32.89 0

T 1.10 55.69 11.3 R

4.11 90.02 15.3 P

4.27 99.91 14.7 Surveillance Weld (Heat # 2721)

V 0.598 70.07 5.8 T

1.10 57.73 7.8 R

4.11 100.31 11.7 P

4.27 96.24 3.9 Note:

(a) Surveillance data from WCAP-14613 [18].

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-1 WCAP-18746-NP October 2022 Revision 1 5

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

The best-estimate copper and nickel weight percent values for the Prairie Island Units 1 and 2 reactor vessel materials are provided in Table 3-1 and Table 3-2 of this report, respectively.

The Position 2.1 chemistry factor calculation is presented in Table 5-1 and Table 5-2 for the Prairie Island Units 1 and 2 surveillance materials, respectively. These values were calculated using the surveillance data summarized in Section 4 of this report. The calculations are performed using the method described in Regulatory Guide 1.99, Revision 2. All of the surveillance weld data considers the chemical composition differences between the surveillance weld and the weld being evaluated, in accordance with the guidance presented at an industry meeting held by the NRC on February 12 and 13, 1998 [11]. In addition to the chemical composition differences, temperature adjustments are considered for the surveillance material from other plants. In this case, the Unit 1 surveillance weld data includes a temperature consideration when applied to the Unit 2 reactor vessel weld. Margin will be applied to the ART calculations in Section 7 and PTS calculations in Appendix D according to the conclusions of the credibility evaluation for the surveillance material, as documented in Section 4.

The Position 1.1 chemistry factors are summarized along with the Position 2.1 chemistry factors in Table 5-3 and Table 5-4 for Prairie Island Units 1 and 2, respectively.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-2 WCAP-18746-NP October 2022 Revision 1 Table 5-1 Calculation of Prairie Island Unit 1 Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule Fluence(a)

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

FF(b)

RTNDT(c)

(°F)

FF*RTNDT

(°F)

FF2 Intermediate Shell Forging C (Tangential)

V 0.609 0.861 31.9 27.5 0.741 P

1.31 1.075 23.2 24.9 1.156 R

4.02 1.357 96.0 130.3 1.842 S

4.39 1.376 101.6 139.8 1.893 N

8.45 1.491 165.5 246.8 2.224 Intermediate Shell Forging C (Axial)

V 0.609 0.861 48.8 42.0 0.741 P

1.31 1.075 33.9 36.4 1.156 R

4.02 1.357 84.0 114.0 1.842 S

4.39 1.376 74.0 101.8 1.893 N

8.45 1.491 147.5 220.0 2.224 SUM:

1083.6 15.715 CF IS Forging C = (FF

RTNDT) ÷ (FF2) = (1083.6) ÷ (15.715) = 69.0°F Surveillance Weld (Heat # 1752)

V 0.609 0.861 34.5(d)

(35.2) 29.7 0.741 P

1.31 1.075 44.7(d)

(45.6) 48.0 1.156 R

4.02 1.357 120.5(d)

(123.0) 163.6 1.842 S

4.39 1.376 157.8(d)

(161.0) 217.1 1.893 N

8.45 1.491 214.6(d)

(219.0) 320.1 2.224 SUM:

778.6 7.857 CF Surv. Weld = (FF

RTNDT) ÷ (FF2) = (778.6) ÷ (7.857) = 99.1°F Notes:

(a) Data taken from Table 4-1.

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

(c) RTNDT values taken from Table 4-1.

(d) The surveillance weld RTNDT values have been adjusted by a factor of 0.98. The calculated adjustment is the ratio of the Position 1.1 CFs of the vessel weld to the surveillance weld (CFVessel Weld / CFSurv. Weld), which is (69.7°F / 70.9°F). The measured (unadjusted) RTNDT values are shown in parenthesis.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-3 WCAP-18746-NP October 2022 Revision 1 Table 5-2 Calculation of Prairie Island Unit 2 Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule Fluence(a)

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

FF(b)

RTNDT(c)

(°F)

FF*RTNDT

(°F)

FF2 Lower Shell Forging D (Axial)

V 0.598 0.856 35.28 30.20 0.733 T

1.10 1.027 29.93 30.73 1.054 R

4.11 1.362 84.73 115.41 1.855 P

4.27 1.370 103.87 142.32 1.877 Lower Shell Forging D (Tangential)

V 0.598 0.856 32.89 28.15 0.733 T

1.10 1.027 55.69 57.17 1.054 R

4.11 1.362 90.02 122.61 1.855 P

4.27 1.370 99.91 136.89 1.877 SUM:

663.5 11.039 CF LS Forging D = (FF

RTNDT) ÷ (FF2) = (663.5) ÷ (11.039) = 60.1°F Surveillance Weld (Heat # 2721)

V 0.598 0.856 80.58(d)

(70.07) 68.98 0.733 T

1.10 1.027 66.39(d)

(57.73) 68.16 1.054 R

4.11 1.362 115.36(d)

(100.31) 157.12 1.855 P

4.27 1.370 110.68(d)

(96.24) 151.65 1.877 SUM:

445.9 5.519 CF Surv. Weld = (FF

RTNDT) ÷ (FF2) = (445.9) ÷ (5.519) = 80.8°F Notes:

(a) Data taken from Table 4-2.

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

(c) RTNDT values taken from Table 4-2.

(d) The surveillance weld RTNDT values have been adjusted by a factor of 1.15. The calculated adjustment is the ratio of the Position 1.1 CFs of the vessel weld to the surveillance weld (CFVessel Weld / CFSurv. Weld), which is (51.6°F / 44.8°F). The measured (unadjusted) RTNDT values are shown in parenthesis.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-4 WCAP-18746-NP October 2022 Revision 1 Table 5-3 Summary of Prairie Island Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material and Identification Number Heat Number (Lot)

Chemistry Factor (°F)

Position 1.1(a)

Position 2.1(b)

Nozzle Shell Forging B 21744/38384 51.0 Intermediate Shell Forging C 21918/38566 44.0 69.0 Lower Shell Forging D 21887/38530 44.0 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 2269 (1180) 79.5 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 1752 (1230) 69.7 99.1 Reactor Vessel Surveillance Material Surveillance Program Weld Metal 1752 (1230) 70.9 Notes:

(a) Position 1.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of this report and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2 [1].

(b) Position 2.1 chemistry factors were taken from Table 5-1 of this report. As discussed in Section 4, both the surveillance forging data and the surveillance weld data were deemed non-credible.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-5 WCAP-18746-NP October 2022 Revision 1 Table 5-4 Summary of Prairie Island Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material and Identification Number Heat Number (Lot)

Chemistry Factor (°F)

Position 1.1(a)

Position 2.1(b)

Upper Shell Forging B 22231/39088 44.0 Intermediate Shell Forging C 22829 44.0 Lower Shell Forging D 22642 51.0 60.1 Upper Shell to Intermediate Shell Circumferential Weld Seam W2 1752 (1263) 69.7 101.4(c)

Intermediate Shell to Lower Shell Circumferential Weld Seam W3 2721 (1263) 51.6 80.8 Reactor Vessel Surveillance Material Prairie Island Unit 2 Surveillance Program Weld Metal 2721 (1263) 44.8 Prairie Island Unit 1 Surveillance Program Weld Metal 1752 (1230) 70.9 Notes:

(a) Position 1.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of this report and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2 [1].

(b) Position 2.1 chemistry factors were taken from Table 5-2 of this report. As discussed in Section 4, the surveillance forging data was deemed non-credible and the surveillance weld data was deemed credible.

(c) Value is determined by recalculating the surveillance weld CF from Table 5-1 and adding the 3°F temperature adjustment before multiplying by the 0.98 CF ratio. This temperature adjustment is meant to account for the Prairie Island Unit 1 surveillance weld material irradiation temperature of 536°F and the time-weighted average inlet temperature for Prairie Island Unit 2 over 54 EFPY of 533°F.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-1 WCAP-18746-NP October 2022 Revision 1 6

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 6.1 OVERALL APPROACH The American Society of Mechanical Engineers (ASME) approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIc, for the metal temperature at that time. KIc is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [3]. The KIc curve is given by the following equation:

)]

(

02

.0

[

734 20 2.

33 NDT RT T

Ic e

K

+

=

(1)

where, KIc (ksiin.)

=

reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This KIc curve is based on the lower bound of static critical KI values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.

6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* KIm + KIt < KIc (2)

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

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-2 WCAP-18746-NP October 2022 Revision 1 For membrane tension, the corresponding KI for the postulated defect is:

)

/

(

Im t

pR M

K i

m

=

(3)

Axial Flaw Methodology where, Mm for an inside axial surface flaw is given by:

Mm

=

1.85 for t < 2, Mm

=

0.926 t for 2

t 3464 Mm

=

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

Mm

=

1.77 for t < 2, Mm

=

0.893 t for 2 464

.3

t Mm

=

3.09 for t > 3.464 Circumferential Flaw Methodology Similarly, Mm for an inside or an outside circumferential surface flaw is given by:

Mm

=

0.89 for t < 2, Mm

=

0.443 t for 2 464

.3

t Mm

=

1.53 for t > 3.464 where, p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).

For bending stress, the corresponding KI for the postulated axial or circumferential defect is:

KIb = Mb

Maximum Bending Stress, where Mb is two-thirds of Mm (4)

The maximum KI produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:

KIt = 0.953 x 10-3 x CR x t2.5 (5) where CR is the cooldown rate in F/hr., or for a postulated axial or circumferential outside surface defect

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-3 WCAP-18746-NP October 2022 Revision 1 KIt = 0.753 x 10-3 x HU x t2.5 (6) where HU is the heatup rate in F/hr.

The through-wall temperature difference associated with the maximum thermal KI can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-2 for the maximum thermal KI:

(a) The maximum thermal KI relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).

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

K C

C C

C a

It =

+

(.

)*

10359 06322 04753 03855 0

1 2

3 (7) or similarly, KIt during heatup for a 1/4-thickness outside axial or circumferential surface defect using the relationship:

a C

C C

C KIt

)

401 0

481 0

630 0

043 1

(

3 2

1 0

+

=

(8) where the coefficients C0, C1, C2, and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

( )

( / )

x C

C x a C x a C x a

=

+

0 1

2 2

3 3

(9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).

Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves [2] Section 2.6 (Equations 2.6.2-4 and 2.6.3-1). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 (Equation 2.6.1-1). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2/hr at 70°F and 0.379 ft2/hr at 550°F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2-°F.

At any time during the heatup or cooldown transient, KIc is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code,Section XI, Paragraph G-2120), the appropriate value for RTNDT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors,

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-4 WCAP-18746-NP October 2022 Revision 1 KIt, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained, and from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable P-T curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the T (temperature) across the vessel wall developed during cooldown results in a higher value of KIc at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIc 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 1/4T location and, therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

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

The third portion of the heatup analysis concerns the calculation of the P-T limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-5 WCAP-18746-NP October 2022 Revision 1 Following the generation of P-T curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20% of the preservice hydrostatic test pressure, which is calculated to be 621 psig. The initial RTNDT values of the reactor vessel closure head and vessel flange are documented in Table 3-3. The limiting unirradiated RTNDT of -4°F for Prairie Island Units 1 and 2 is associated with the vessel flange of the Prairie Island Unit 1 vessel, so the minimum allowable temperature of this region is 116°F at pressures greater than 621 psig without margins for instrument uncertainties. This limit is shown in Figure 8-1 and Figure 8-2.

6.4 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded. It is determined by the highest reference temperature, RTNDT, in the closure flange region. This requirement is established in Appendix G to 10 CFR 50 [4]. Per the NRC-approved methodology in WCAP-14040-A, Revision 4 [2], the minimum boltup temperature should be 60°F or the limiting unirradiated RTNDT of the closure flange region, whichever is higher. Since the limiting unirradiated RTNDT of this region is below 60F per Table 3-3, the minimum boltup temperature for the Prairie Island Units 1 and 2 reactor vessels is 60°F without margins for instrument uncertainties. This limit is shown in Figure 8-1 and Figure 8-2.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-1 WCAP-18746-NP October 2022 Revision 1 7

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2 [1], the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + RTNDT + Margin (10)

Initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [8]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided if there are sufficient test results to establish a mean and standard deviation for the class.

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)

(11)

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) = fsurface

e (-0.24x)

(12) where x inches (reactor vessel cylindrical shell beltline thickness is 6.692 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 11 to calculate the RTNDT at the specific depth.

The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-18124-NP-A.

Table 7-1 and Table 7-2 contain the surface fluence values at 54 EFPY for Prairie Island Units 1 and 2, respectively, which were used for the development of the P-T limit curves contained in this report. Table 7-1 and Table 7-2 also contain 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 54 EFPY ART values for the Prairie Island Units 1 and 2 reactor vessel materials.

Margin is calculated as M = 2 2

2

+

I

. The standard deviation for the initial RTNDT margin term (I) is 0F when the initial RTNDT is a measured value, and 17F when a generic value is available. The standard deviation for the RTNDT margin term,, is 17F for plates or forgings when surveillance data is not used or is non-credible, and 8.5F (half the value) for plates or forgings when credible surveillance data is used.

For welds, is equal to 28F when surveillance capsule data is not used or is non-credible and is 14F (half the value) when credible surveillance capsule data is used. The value for need not exceed 0.5 times the mean value of RTNDT.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-2 WCAP-18746-NP October 2022 Revision 1 Contained in Table 7-3 and Table 7-4 are the 54 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Prairie Island Unit 1 heatup and cooldown curves. Contained in Table 7-5 and Table 7-6 are the 54 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Prairie Island Unit 2 heatup and cooldown curves.

The inlet/outlet nozzle forging weld materials for Prairie Island Units 1 and 2 have projected fluence values that do not exceed the 1 x 1017 n/cm2 fluence threshold of RIS 2014-11 [6] at 54 EFPY per Table 2-8 (Unit 1) and Table 2-17 (Unit 2). Note that neither Table 2-8 nor Table 2-17 provide fluence values for the outlet nozzle, which are at a higher elevation than the inlet nozzles. The projected fluence value for the inlet nozzle forging weld material provides a conservative estimate of the fluence values of the outlet nozzles. Therefore, neutron radiation embrittlement need not be considered herein for the nozzle forging or weld materials.

Thus, ART calculations for the inlet and outlet nozzle forging and weld materials are excluded from Table 7-3 through Table 7-6.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-3 WCAP-18746-NP October 2022 Revision 1 Table 7-1 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 1 Reactor Vessel at 54 EFPY Reactor Vessel Material Surface Fluence(a)

(n/cm2, E > 1.0 MeV)

Surface FF(b) 1/4T Fluence(a)

(n/cm2, E > 1.0 MeV) 1/4T FF(b) 3/4T Fluence(a)

(n/cm2, E > 1.0 MeV) 3/4T FF(b)

Nozzle Shell Forging B 3.37 1.318 2.26 1.220 1.01 1.003 Intermediate Shell Forging C 5.63 1.425 3.77 1.343 1.69 1.144 Lower Shell Forging D 5.53 1.422 3.70 1.339 1.66 1.139 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 3.63 1.335 2.43 1.239 1.09 1.024 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 5.53 1.422 3.70 1.339 1.66 1.139 Notes:

(a) 54 EFPY fluence values are documented in Table 2-8. 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (6.692 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).

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

Table 7-2 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 2 Reactor Vessel at 54 EFPY Reactor Vessel Material Surface Fluence(a)

(n/cm2, E > 1.0 MeV)

Surface FF(b) 1/4T Fluence(a)

(n/cm2, E > 1.0 MeV) 1/4T FF(b) 3/4T Fluence(a)

(n/cm2, E > 1.0 MeV) 3/4T FF(b)

Upper Shell Forging B 3.37 1.318 2.26 1.220 1.01 1.003 Intermediate Shell Forging C 5.66 1.426 3.79 1.344 1.70 1.146 Lower Shell Forging D 5.58 1.423 3.73 1.341 1.67 1.142 Upper Shell to Intermediate Shell Circumferential Weld - Seam W2 3.61 1.334 2.42 1.238 1.08 1.022 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 5.58 1.423 3.73 1.341 1.67 1.142 Notes:

(a) 54 EFPY fluence values are documented in Table 2-17. 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (6.692 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).

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

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-4 WCAP-18746-NP October 2022 Revision 1 Table 7-3 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)

(°F) 1/4T Fluence(b)

(n/cm2, E > 1.0 MeV) 1/4T FF(b)

RTNDT(U)(c)

(°F)

RTNDT

(°F)

I

(°F)

(d)

(°F)

Margin

(°F)

ART

(°F)

Nozzle Shell Forging B 1.1 51.0 2.26 1.220

-4 62.2 0

17.0 34.0 92.2 Intermediate Shell Forging C 1.1 44.0 3.77 1.343 14 59.1 0

17.0 34.0 107.1 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 69.0 3.77 1.343 14 92.7 0

17.0 34.0 140.7 Lower Shell Forging D 1.1 44.0 3.70 1.339

-4 58.9 0

17.0 34.0 88.9 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 1.1 79.5 2.43 1.239 0

98.5 17 28.0 65.5 164.0 Intermediate to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 1.1 69.7 3.70 1.339

-13 93.3 0

28.0 56.0 136.3 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 99.1 3.70 1.339

-13 132.7 0

28.0 56.0 175.7 Notes:

(a) Data is from Table 5-3.

(b) Data is from Table 7-1.

(c) Data is from Table 3-1.

(d) As discussed in Section 4, the intermediate shell forging material surveillance data and the weld Heat # 1752 surveillance data were both determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or 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 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-5 WCAP-18746-NP October 2022 Revision 1 Table 7-4 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)

(°F) 3/4T Fluence(b)

(n/cm2, E > 1.0 MeV) 3/4T FF(b)

RTNDT(U)(c)

(°F)

RTNDT

(°F)

I

(°F)

(d)

(°F)

Margin

(°F)

ART

(°F)

Nozzle Shell Forging B 1.1 51.0 1.01 1.003

-4 51.1 0

17.0 34.0 81.1 Intermediate Shell Forging C 1.1 44.0 1.69 1.144 14 50.3 0

17.0 34.0 98.3 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 69.0 1.69 1.144 14 78.9 0

17.0 34.0 126.9 Lower Shell Forging D 1.1 44.0 1.66 1.139

-4 50.1 0

17.0 34.0 80.1 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 1.1 79.5 1.09 1.024 0

81.4 17 28.0 65.5 146.9 Intermediate to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 1.1 69.7 1.66 1.139

-13 79.4 0

28.0 56.0 122.4 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 99.1 1.66 1.139

-13 112.9 0

28.0 56.0 155.9 Notes:

(a) Data is from Table 5-3.

(b) Data is from Table 7-1.

(c) Data is from Table 3-1.

(d) As discussed in Section 4, the intermediate shell forging material surveillance data and the weld Heat # 1752 surveillance data were both determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or 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 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-6 WCAP-18746-NP October 2022 Revision 1 Table 7-5 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)

(°F) 1/4T Fluence(b)

(n/cm2, E > 1.0 MeV) 1/4T FF(b)

RTNDT(U)(c)

(°F)

RTNDT

(°F)

I

(°F)

(d)

(°F)

Margin

(°F)

ART

(°F)

Upper Shell Forging B 1.1 44.0 2.26 1.220

-13 53.7 0

17.0 34.0 74.7 Intermediate Shell Forging C 1.1 44.0 3.79 1.344 14 59.2 0

17.0 34.0 107.2 Lower Shell Forging D 1.1 51.0 3.73 1.341

-4 68.4 0

17.0 34.0 98.4 Using Non-credible Prairie Island Unit 2 Surveillance Data 2.1 60.1 3.73 1.341

-4 80.6 0

17.0 34.0 110.6 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 1.1 69.7 2.42 1.238

-13 86.3 0

28.0 56.0 129.3 Using Non-credible Prairie Island Unit 1 Surveillance Data (Heat # 1752) 2.1 101.4 2.42 1.238

-13 125.5 0

28.0 56.0 168.5 Intermediate to Lower Shell Circumferential Weld - Seam W3 1.1 51.6 3.73 1.341

-31 69.2 0

28.0 56.0 94.2 Using Credible Prairie Island Unit 2 Surveillance Data (Heat # 2721) 2.1 80.8 3.73 1.341

-31 108.4 0

14.0 28.0 105.4 Notes:

(a) Data is from Table 5-4.

(b) Data is from Table 7-2.

(c) Data is from Table 3-2.

(d) As discussed in Section 4, the lower shell forging material surveillance data was determined to be non-credible, while the weld Heat # 2721 surveillance data was determined to be credible. In addition, as discussed in Section 4, the weld Heat # 1752 surveillance data was determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1],

the base metal = 17°F for Position 1.1 or 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 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-7 WCAP-18746-NP October 2022 Revision 1 Table 7-6 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)

(°F) 3/4T Fluence(b)

(n/cm2, E > 1.0 MeV) 3/4T FF(b)

RTNDT(U)(c)

(°F)

RTNDT

(°F)

I

(°F)

(d)

(°F)

Margin

(°F)

ART

(°F)

Upper Shell Forging B 1.1 44.0 1.01 1.003

-13 44.1 0

17.0 34.0 65.1 Intermediate Shell Forging C 1.1 44.0 1.70 1.146 14 50.4 0

17.0 34.0 98.4 Lower Shell Forging D 1.1 51.0 1.67 1.142

-4 58.2 0

17.0 34.0 88.2 Using Non-credible Prairie Island Unit 2 Surveillance Data 2.1 60.1 1.67 1.142

-4 68.8 0

17.0 34.0 98.6 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 1.1 69.7 1.08 1.022

-13 71.2 0

28.0 56.0 114.2 Using Non-credible Prairie Island Unit 1 Surveillance Data (Heat # 1752) 2.1 101.4 1.08 1.022

-13 103.6 0

28.0 56.0 146.6 Intermediate to Lower Shell Circumferential Weld - Seam W3 1.1 51.6 1.67 1.142

-31 58.9 0

28.0 56.0 83.9 Using Credible Prairie Island Unit 2 Surveillance Data (Heat # 2721) 2.1 80.8 1.67 1.142

-31 92.3 0

14.0 28.0 89.3 Notes:

(a) Data is from Table 5-4.

(b) Data is from Table 7-2.

(c) Data is from Table 3-2.

(d) As discussed in Section 4, the lower shell forging material surveillance data was determined to be non-credible, while the weld Heat # 2721 surveillance data was determined to be credible. In addition, as discussed in Section 4, the weld Heat # 1752 surveillance data was determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1],

the base metal = 17°F for Position 1.1 or 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 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-1 WCAP-18746-NP October 2022 Revision 1 8

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES P-T limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 6 and 7 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4 [2].

The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figure 8-1 through Figure 8-2. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed in the following paragraphs.

The reactor must not be made critical until P-T combinations are to the right of the criticality limit line shown in Figure 8-1 (heatup curve only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G as follows:

1.5 KIm < KIc (13)

where, KIm is the stress intensity factor covered by membrane (pressure) stress [see Equation (3) in Section 6],

KIc = 33.2 + 20.734 e [0.02 (T - RTNDT)] [see Equation (1) in Section 6],

T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies P-T limits for core operation in order to provide additional margin during actual power production. The P-T limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding P-T curve for heatup and cooldown calculated as described in Section 6 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the inservice hydrostatic leak tests for the Prairie Island Units 1 and 2 reactor vessels at 54 EFPY is 219F without uncertainties. This temperature is the minimum permissible temperature at which design pressure can be reached during a hydrostatic test per Equation (13). The vertical line drawn from these points on the P-T curve, intersecting a curve 40°F higher than the P-T limit curve, constitutes the limit for core operation for the reactor vessel.

The ART values were calculated in Table 7-3 through Table 7-6. The limiting ART values for Prairie Island Units 1 and 2 are for Unit 1 Intermediate Shell to Lower Shell Circumferential Weld Seam W3 (Heat

1752). However, since this material is a Circumferential Flaw material, the applied membrane

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-2 WCAP-18746-NP October 2022 Revision 1 (pressure) stress and resulting stress intensity factor at the postulated flaw location are much lower than for the most limiting Axial Flaw material. Consequently, this material may not produce the most limiting P-T limit curves. This is illustrated in the figures in Section 9. The ART values considered for circumferential flaws in the generation of the 54 EFPY P-T limit curves are provided in Table 8-1, and bound the ART values in Table 7-3 through Table 7-6. Therefore, the most limiting material with an axially oriented flaw must also be considered. The limiting material with an axially oriented flaw is Unit 1 Intermediate Shell Forging C. It is noted that the 1/4T and 3/4T ART values for Unit 1 Intermediate Shell Forging C at 54 EFPY (140.7°F and 126.9°F, respectively) are below those used to generate the P-T curves in WCAP-14780 and currently implemented in the PTLR (154°F and 136°F, respectively); thereby, validating the PTLR curve for the material with a postulated axial flaw. However, the ART values for Unit 1 Intermediate Shell Forging C were conservatively increased to account for future perturbations such as the Unit 2 Surveillance Capsule N results. The ART values, considered in the P-T limit curves development are summarized in Table 8-1.

Table 8-1 Summary of the ART Values Used in the Generation of the Prairie Island Units 1 and 2 Heatup and Cooldown Curves at 54 EFPY Flaw Orientation 1/4T Limiting ART 3/4T Limiting ART Axial 170°F 160°F Unit 1 Intermediate Shell Forging C Circumferential 179°F 159°F Unit 1 Intermediate Shell to Lower Shell Circumferential Weld Seam W3 (Heat # 1752)

Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. Figure 8-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, -20, -40, -60, and -100°F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. The data points used for developing the heatup and cooldown P-T limit curves shown in Figure 8-1 and Figure 8-2 are presented in Table 8-2 and Table 8-3. Vacuum refill limits for the Reactor Coolant System (RCS) are displayed on Figure 8-1 and Figure 8-2 by showing a minimum pressure of 0 psia.

Inlet and Outlet Nozzles P-T Limit Curves NRC Regulatory Issue Summary (RIS) 2014-11 [6] 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 [16] 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) bound the generic nozzle P-T limit curves. The

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-3 WCAP-18746-NP October 2022 Revision 1 results and conclusions of PWROG-15109-NP-A are applicable as long as the plant-specific Prairie Island Units 1 and 2 fluence of the nozzle corners remains less than the screening criterion of 4.28 x 1017 n/cm2, as described in PWROG-15109-NP-A. Table 2-8 and Table 2-17 demonstrate Prairie Island Units 1 and 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 Prairie Island Units 1 and 2. Therefore, the concerns the concerns of RIS 2014-11 are adequately addressed.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-4 WCAP-18746-NP October 2022 Revision 1 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Unit 1, Intermediate Shell Forging C LIMITING ART VALUES AT 54 EFPY:

1/4T, 170F (Axial Flaw) 3/4T, 160F (Axial Flaw)

Figure 8-1 Prairie Island Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100oF/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/KIc)

-250 0

250 500 750 1000 1250 1500 1750 2000 2250 2500 0

50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (PSIG)

Moderator Temperature (Deg. F)

OperlimAnalysis Version:5.4 Run:27388 Operlim.xlsm Version: 5.4.1 Unacceptable Operation Acceptable Operation Criticality Limit based on inservice hydrostatic test temperature (219ºF) for the service period up to 54 EFPY Heatup Rate 60 F/Hr Heatup Rate 100 F/Hr Critical Limit 60 F/Hr Critical Limit 100 F/Hr Leak Test Limit Boltup Temp.

Prairie Island Unit 1 54 EFPY Curves using K1c, Axial Flaw, No instrumentation errors, ART 170_160, No Flange Notch Lower Limit for RCS pressure is 0 psia

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-5 WCAP-18746-NP October 2022 Revision 1 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Unit 1, Intermediate Shell Forging C LIMITING ART VALUES AT 54 EFPY:

1/4T, 170F (Axial Flaw) 3/4T, 160F (Axial Flaw)

Figure 8-2 Prairie Island Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)

-250 0

250 500 750 1000 1250 1500 1750 2000 2250 2500 0

50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (PSIG)

Moderator Temperature (Deg. F)

OperlimAnalysis Version:5.4 Run:27388 Operlim.xlsm Version: 5.4.1 Unacceptable Operation Acceptable Operation Cooldown Rates 0 F/hr

-20 F/hr

-40 F/hr

-60 F/hr

-100 F/hr Prairie Island Unit 1 54 EFPY Curves using K1c, Axial Flaw, No instrumentation errors, ART 170_160, No Flange Notch Steady State and Cooldown Curves Lower Limit for RCS pressure is 0 psia

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-6 WCAP-18746-NP October 2022 Revision 1 Table 8-2 Prairie Island Units 1 and 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 60

-14.7 219

-14.7 60

-14.7 219

-14.7 60 621 219 1101 60 621 219 954 65 621 220 1111 65 621 220 961 70 621 225 1159 70 621 225 998 75 621 230 1212 75 621 230 1039 80 621 235 1270 80 621 235 1084 85 621 240 1335 85 621 240 1134 90 621 245 1406 90 621 245 1189 95 621 250 1485 95 621 250 1250 100 621 255 1572 100 621 255 1317 105 621 260 1668 105 621 260 1391 110 621 265 1774 110 621 265 1472 115 621 270 1892 115 621 270 1563 116 621 275 2021 116 621 275 1662 116 783 280 2164 116 721 280 1772 120 793 285 2322 120 727 285 1894 125 807 125 736 290 2028 130 823 130 747 295 2175 135 840 135 759 300 2338 140 859 140 772 145 881 145 788 150 905 150 805 155 931 155 825 160 960 160 847 165 992 165 871 170 1028 170 898 175 1067 175 928 180 1111 180 961 185 1159 185 998 190 1212 190 1039 195 1270 195 1084 200 1335 200 1134 205 1406 205 1189 210 1485 210 1250 215 1572 215 1317

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-7 WCAP-18746-NP October 2022 Revision 1 Table 8-2 Prairie Island Units 1 and 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 220 1668 220 1391 225 1774 225 1472 230 1892 230 1563 235 2021 235 1662 240 2164 240 1772 245 2322 245 1894 250 2028 255 2175 260 2338 Leak Test Limit T (°F)

P (psig) 200 2000 219 2485

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-8 WCAP-18746-NP October 2022 Revision 1 Table 8-3 Prairie Island Units 1 and 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 60

-14.7 60

-14.7 60 621 60 621 60 621 60 621 60 601 65 621 65 621 65 621 65 621 65 607 70 621 70 621 70 621 70 621 70 614 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 116 621 116 621 116 621 116 621 116 621 116 849 116 825 116 800 116 777 116 729 120 861 120 838 120 814 120 791 120 745 125 878 125 856 125 833 125 811 125 768 130 897 130 875 130 854 130 833 130 793 135 918 135 897 135 877 135 857 135 820 140 941 140 921 140 903 140 885 140 851 145 966 145 948 145 931 145 915 145 885 150 994 150 978 150 962 150 948 150 922 155 1025 155 1010 155 997 155 985 155 964 160 1059 160 1046 160 1035 160 1026 160 1010 165 1096 165 1086 165 1078 165 1071 165 1061 170 1138 170 1131 170 1125 170 1121 170 1118 175 1184 175 1179 175 1177 175 1176 175 1176 180 1235 180 1233 180 1233 180 1233 180 1233 185 1291 185 1291 185 1291 185 1291 185 1291 190 1353 190 1353 190 1353 190 1353 190 1353 195 1422 195 1422 195 1422 195 1422 195 1422 200 1498 200 1498 200 1498 200 1498 200 1498

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-9 WCAP-18746-NP October 2022 Revision 1 Table 8-3 Prairie Island Units 1 and 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 205 1581 205 1581 205 1581 205 1581 205 1581 210 1674 210 1674 210 1674 210 1674 210 1674 215 1777 215 1777 215 1777 215 1777 215 1777 220 1890 220 1890 220 1890 220 1890 220 1890 225 2015 225 2015 225 2015 225 2015 225 2015 230 2153 230 2153 230 2153 230 2153 230 2153 235 2306 235 2306 235 2306 235 2306 235 2306 240 2475 240 2475 240 2475 240 2475 240 2475

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-1 WCAP-18746-NP October 2022 Revision 1 9

HEATUP AMD COOLDOWN LIMITS APPLICABILITY AND MARGIN ASSESSMENT This section provides a comparison of the heatup and cooldown P-T limit curves currently implemented in the Prairie Island Units 1 and 2 PTLR and the heatup and cooldown P-T limit curves generated in this report.

The curves contained in the PTLR were generated in WCAP-14780 [9] and were generated consistent with WCAP-14040-A, Revision 2. WCAP-14780 used KIa stress intensity factors and an Axial-Flaw methodology even though the limiting materials are circumferential welds. KIa is a toughness based on the lower bound of crack arrest toughness, whereas KIc is a toughness based on the lower bound of crack initiation toughness. This was consistent with the requirements of Revision 2 of WCAP-14040-NP-A.

Since the time the curves from WCAP-14780 were generated, WCAP-14040-A has been revised to Revision 4 [2] and approved for general use by the NRC. As discussed in Section 6, WCAP-14040-A, Revision 4 utilizes the 1998 version of the ASME Code through the Summer 2000 Addenda, which allows the use of the less restrictive KIc stress intensity factors and allows the use of the less restrictive Circ-Flaw methodology (formerly known as ASME Code Cases N-640 and N-588, respectively). Because less restrictive methodologies are used in this report; the current PTLR curves may remain valid, despite an increase in the maximum ART values.

The curves generated in this report are compared to the P-T limit curves from WCAP-14780 in Figure 9-1 and Figure 9-2. The curves generated with an axial flaw and KIc are shown as dashed blue lines. The curves generated with a circumferential flaw and KIc are shown as orange dotted lines. The curves from WCAP-14780 are shown as solid green lines. Note that the curves do not have margins for instrumentation errors, but they do have the Appendix G flange notch requirements included with the exception of the circumferential flaw curves which do not have the flange notch included for visual clarity.

In the flange region, the minimum pressure difference, at a constant temperature, between the current P-T limit curves and the new curves developed in this report is 0 psid for the cooldown curves. This is driven by the flange notch requirements of 10 CFR 50, Appendix G [4] being identical for each curve. The pressure margins at temperatures past the flange notch region show greater margin, with the minimum margin being 158 psid at a temperature of 116°F during the 100°F/hr heatup.

P-T Limit Curve Applicability Conclusion As shown in Figure 9-1 and Figure 9-2, the margins between the curves developed in this report and the current P-T limit curves illustrate that the current P-T limit curves remain applicable through 54 EPFY.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-2 WCAP-18746-NP October 2022 Revision 1 Figure 9-1 Heatup Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws

-250 0

250 500 750 1000 1250 1500 1750 2000 2250 2500 0

50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (psig)

Moderator Temperature

(°F)

Axial Flaw - ART Values (170, 160)

PTLR - ART Values (154, 136)

Circ Flaw - ART Values (179, 159)

Acceptable Operation Unacceptable Operation

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Westinghouse Non-Proprietary Class 3 9-3 WCAP-18746-NP October 2022 Revision 1 Figure 9-2 Cooldown Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws

-250 0

250 500 750 1000 1250 1500 1750 2000 2250 2500 0

50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (psig)

Moderator Temperature

(°F)

Axial Flaw - ART Values (170, 160)

PTLR - ART Values (154, 136)

Circ Flaw - ART Values (179, 159)

Acceptable Operation Unacceptable Operation

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 10-1 WCAP-18746-NP October 2022 Revision 1 10 REFERENCES

1.

U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988. [Agencywide Document Management System (ADAMS) Accession Number ML003740284]

2.

Westinghouse Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004. [ADAMS Accession Number ML050120209]

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.

Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

5.

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, July 1999.

6.

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 Number ML14149A165]

7.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, Fracture Toughness Requirements, Revision 4, U.S. Nuclear Regulatory Commission, March 2019. [ADAMS Accession Number ML18338A516]

8.

ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Class 1 Components.

9.

Westinghouse Report WCAP-14780, Revision 3, Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, February 1998.

10. 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.

11. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

[ADAMS Accession Number ML110070570]

12. ASTM E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results, 2018.

13. ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706 (ID), 1994.

14. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [ADAMS Accession Number ML010890301]

15. Westinghouse Report WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018. [ADAMS Accession Number ML18204A010]

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 10-2 WCAP-18746-NP October 2022 Revision 1

16. Pressurized Water Reactor Owners Group (PWROG) Report PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020. [ADAMS Accession Number ML20024E573]

17. Westinghouse Report WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, November 2021.

18. Westinghouse Report WCAP-14613, Revision 2, Analysis of Capsule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, February 1998.

19. Westinghouse Report WCAP-18124-NP-A, Revision 0, Supplement 1-P, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, December 2020. [ADAMS Accession Number ML20344A388]

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-1 WCAP-18746-NP October 2022 Revision 1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt)

Table A-1 and Table A-2 contain the thermal stress intensity factors (KIt) for the maximum heatup and cooldown rates at 54 EFPY for Prairie Island Units 1 and 2. The reactor vessel cylindrical shell radii to the 1/4T and 3/4T locations are as follows:

1/4T Radius = 67.869 inches 3/4T Radius = 71.215 inches

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Westinghouse Non-Proprietary Class 3 A-2 WCAP-18746-NP October 2022 Revision 1 Table A-1 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Heatup Curves Water Temp.

(F)

Vessel Temperature at 1/4T Location for 100F/hr Heatup (F) 1/4T Thermal Stress Intensity Factor (ksi in.)

Vessel Temperature at 3/4T Location for 100F/hr Heatup (F) 3/4T Thermal Stress Intensity Factor (ksi in.)

60 56.404

-0.958 55.128 0.520 65 59.579

-2.202 55.759 1.442 70 63.008

-3.125 57.148 2.208 75 66.730

-3.950 59.188 2.848 80 70.703

-4.586 61.743 3.369 85 74.816

-5.138 64.738 3.800 90 79.119

-5.574 68.083 4.154 95 83.508

-5.952 71.717 4.448 100 88.030

-6.253 75.584 4.691 105 92.606

-6.518 79.638 4.897 110 97.271

-6.730 83.845 5.068 115 101.971

-6.919 88.178 5.215 120 106.730

-7.072 92.613 5.338 125 111.513

-7.211 97.131 5.445 130 116.335

-7.324 101.719 5.536 135 121.173

-7.429 106.362 5.616 140 126.037

-7.515 111.050 5.686 145 130.912

-7.597 115.777 5.748 150 135.804

-7.666 120.534 5.803 155 140.704

-7.733 125.316 5.854 160 145.615

-7.790 130.119 5.899 165 150.532

-7.846 134.938 5.942 170 155.456

-7.896 139.772 5.981 175 160.384

-7.945 144.617 6.018 180 165.316

-7.989 149.471 6.052 185 170.252

-8.033 154.333 6.086 190 175.190

-8.074 159.201 6.117 195 180.131

-8.115 164.074 6.148 200 185.072

-8.153 168.951 6.178 205 190.017

-8.193 173.832 6.207 210 194.961

-8.229 178.716 6.236

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Westinghouse Non-Proprietary Class 3 A-3 WCAP-18746-NP October 2022 Revision 1 Table A-2 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Cooldown Curves Water Temp.

(F)

Vessel Temperature at 1/4T Location for 100F/hr Cooldown (F) 100F/hr Cooldown 1/4T Thermal Stress Intensity Factor (ksi in.)

210 226.075 8.848 205 221.021 8.811 200 215.968 8.774 195 210.914 8.737 190 205.861 8.701 185 200.808 8.664 180 195.755 8.627 175 190.702 8.591 170 185.649 8.554 165 180.596 8.518 160 175.543 8.482 155 170.491 8.445 150 165.438 8.409 145 160.386 8.373 140 155.334 8.337 135 150.281 8.301 130 145.229 8.265 125 140.177 8.229 120 135.126 8.194 115 130.074 8.158 110 125.022 8.123 105 119.970 8.087 100 114.919 8.052 95 109.867 8.017 90 104.816 7.982 85 99.765 7.947 80 94.714 7.912 75 89.663 7.877 70 84.612 7.842 65 79.561 7.807 60 74.512 7.772

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-1 WCAP-18746-NP October 2022 Revision 1 APPENDIX B OTHER RCPB FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [B-1], requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature requirement (LST) for all RCPB components, which is specified in NB-3210 and NB-2332(b) of the Section III ASME Code [B-2], is the relevant requirement that would affect the P-T limits. This requirement is applicable to ferritic materials outside of the reactor vessel with a nominal wall thickness greater than 2 1/2 inches, such as piping, pumps and valves. The Prairie Island Units 1 and 2 reactor coolant system components do not contain ferritic materials in the Class 1 piping, pumps and valves. Therefore, the LST requirements of NB-2332(b) and NB-3210 are not applicable to the Prairie Island Units 1 and 2 P-T limits.

The other ferritic RCPB components that are not part of the reactor vessel beltline consist of the reactor vessel closure head, pressurizer, and steam generators.

RIS 2014-11 also addresses other ferritic components of the reactor coolant system relative to P-T limits, and states the following:

As specified in Sections I and IV.A of 10 CFR Part 50, Appendix G, ferritic RCPB components outside of the reactor vessel must meet the applicable requirements of ASME Code,Section III, Rules for Construction of Nuclear Facility Components.

The reactor vessel closure head flange materials have been considered in the development of P-T limits, see Section 6.3 of this report for further detail. Furthermore, the replacement reactor vessel closure heads were constructed to the 1998 Edition through 2000 Addenda of Section III of the ASME Code and has met all applicable requirements at the time of construction. The steam generators were constructed to the 1995 Edition through 1996 Addenda of Section III of the ASME Code and have met all applicable requirements.

The Unit 1 pressurizer was constructed to the 1965 Edition through 1966 Summer Addenda of Section III of the ASME Code, and the Unit 2 pressurizer was constructed to the 1965 Edition through 1966 Winter Addenda of Section III of the ASME Code. These Prairie Island Units 1 and 2 primary system components are analyzed to the identified ASME Code Section III Editions and met all applicable requirements at the time of construction. In addition, these components have not undergone neutron embrittlement. Therefore, no further consideration is necessary for these components with regard to P-T limits.

B.1 REFERENCES B-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

B-2 ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Class 1 Components.

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Westinghouse Non-Proprietary Class 3 C-1 WCAP-18746-NP October 2022 Revision 1 APPENDIX C UPPER-SHELF ENERGY EVALUATION Charpy upper-shelf energy (USE) is associated with the determination of acceptable reactor pressure vessel (RPV) toughness during the licensed operating period.

The requirements on USE are included in 10 CFR 50, Appendix G [C-1]. 10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the USE of any 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 predict the decrease in USE with irradiation, depending on the availability of credible surveillance capsule data as defined in Regulatory Guide 1.99, Revision 2 [C-2].

For vessel materials that are not in the surveillance program or have non-credible data, the Charpy USE (Position 1.2) is assumed to decrease as a function of fluence and copper content, as indicated in Figure 2 of Regulatory Guide 1.99, Revision 2.

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 to predict the change in USE (Position 2.2) of the RPV material due to irradiation by plotting the reduced plant surveillance data on Figure 2 of the Guide (Figure C-1 and Figure C-2 of this report for Prairie Island Units 1 and 2, respectively) and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.

The 54 EFPY (EOLE) Position 1.2 USE values of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2.

The predicted Position 2.2 USE values are determined for the reactor vessel materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding 1/4T fluence projection.

The projected USE values were calculated to determine if the Prairie Island Units 1 and 2 beltline materials remain above the 50 ft-lb criterion at 54 EFPY (EOLE). These calculations are summarized in Table C-1 and Table C-2 for Prairie Island Units 1 and 2, respectively. Note, even though some surveillance data for both units is deemed non-credible, it is still used in accordance with Regulatory Guide 1.99, Revision 2, Position 2.2. The data from the surveillance materials are determined to be non-credible by Credibility Criterion 3. Credibility Criterion 3 indicates that even if the surveillance data are not considered credible for determination of RTNDT, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82.

USE Conclusion For Prairie Island Units 1 and 2, all of the beltline materials are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G) through EOLE (54 EFPY) with the exception of the Unit 1 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 without using surveillance data (Position 1.2). However, taking into account the surveillance data (Position 2.2), which should be used in preference to Position 1.2 per Regulatory Guide 1.99, Revision 2, the USE value for the Intermediate

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Westinghouse Non-Proprietary Class 3 C-2 WCAP-18746-NP October 2022 Revision 1 Shell to Lower Shell Circumferential Weld - Seam W3 is projected to remain above the 50 ft-lb screening criterion value through EOLE. Therefore, all of the Prairie Island Units 1 and 2 reactor vessel materials are projected to remain above the 10 CFR 50, Appendix USE screening criterion value of 50 ft-lb at EOLE (54 EFPY).

C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

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

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-3 WCAP-18746-NP October 2022 Revision 1 Table C-1 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 1 Beltline Materials Reactor Vessel Material Wt %

Cu(a)

EOLE 1/4T Fluence(b)

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

Initial USE(a)

(ft-lb)

Projected USE Decrease (%)

Projected EOLE USE (ft-lb)

Position 1.2(c)

Nozzle Shell Forging B 0.08 2.26 84 23 65 Intermediate Shell Forging C 0.07 3.77 143 26 106 Lower Shell Forging D 0.07 3.70 88 26 65 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 0.15 2.43 84 36 54 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 0.13 3.70 78.5 38 49 Position 2.2(d)

Intermediate Shell Forging C 3.77 143 14 123 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 3.70 78.5 9(e) 71 Notes:

(a)

Data taken from Table 3-1 of this report. If the base metal and weld Cu weight percentages are below the minimum value presented in Figure 2 of [C-2] (0.1 for base metal and 0.05 for welds), then the Cu weight percentages were conservatively rounded up to the minimum value.

(b)

Values taken from Table 7-1. Fluence values above 1017 n/cm2 (E > 1.0 MeV) but below 2 x 1017 n/cm2 (E > 1.0 MeV) were rounded to 2 x 1017 n/cm2 (E > 1.0 MeV) when determining the % decrease because 2 x 1017 n/cm2 is the lowest fluence displayed in Figure 2 of Regulatory Guide 1.99, Revision 2 [C-2].

(c)

Percentage USE decrease values are based on Position 1.2 of Regulatory Guide 1.99, Revision 2 [C-2] and were calculated by plotting the 1/4T fluence values on Figure 2 of the Regulatory Guide and using the material-specific Cu wt. % values. The percent-loss lines were extended into the low fluence area of Regulatory Guide 1.99, Revision 2, Figure 2, i.e., below 1018 n/cm2, in order to determine the USE % decrease as needed.

(d)

Percentage USE decrease values are based on Position 2.2 of [C-2]. Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of [C-2]) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.

(e)

This Regulatory Guide 1.99, Revision 2, Position 2.2 USE %-decrease determination is based on a Capsule R Heat # 1752 surveillance weld %-decrease of 9%, which is based on an unirradiated USE of 82.5 ft-lb compared to the unirradiated USE for the reactor vessel Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) unirradiated USE of 78.5 ft-lb.

The 82.5 ft-lb value considers an additional data point with 97% shear. Whereas the reactor vessel weld unirradiated USE value of 78.5 ft-lb is consistent with FSAR Table 4.7-2 and only considers data with 100% shear. Each value is conservative for their respective use. A higher unirradiated USE value, when evaluating surveillance data, will result in a higher measured %-decrease, and a lower unirradiated USE value, when performing USE projection, will result in a lower USE projection.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-4 WCAP-18746-NP October 2022 Revision 1 Table C-2 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 2 Beltline Materials Reactor Vessel Material Wt %

Cu(a)

EOLE 1/4T Fluence(b)

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

Initial USE(a)

(ft-lb)

Projected USE Decrease (%)

Projected EOLE USE (ft-lb)

Position 1.2(c)

Upper Shell Forging B 0.07 2.26 85 23 65 Intermediate Shell Forging C 0.07 3.79 112 26 83 Lower Shell Forging D 0.08 3.73 108 26 80 Upper Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 1752) 0.13 2.42 78.5 33 53 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 2721) 0.09 3.73 103 32 70 Position 2.2(d)

Lower Shell Forging D 3.73 108 19 87 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 2721) 3.73 103 12 91 Notes:

(a)

Data taken from Table 3-2 of this report. If the base metal and weld Cu weight percentages are below the minimum value presented in Figure 2 of [C-2] (0.1 for base metal and 0.05 for welds), then the Cu weight percentages were conservatively rounded up to the minimum value.

(b)

Values taken from Table 7-2. Fluence values above 1017 n/cm2 (E > 1.0 MeV) but below 2 x 1017 n/cm2 (E > 1.0 MeV) were rounded to 2 x 1017 n/cm2 (E > 1.0 MeV) when determining the % decrease because 2 x 1017 n/cm2 is the lowest fluence displayed in Figure 2 of Regulatory Guide 1.99, Revision 2 [C-2].

(c)

Percentage USE decrease values are based on Position 1.2 of Regulatory Guide 1.99, Revision 2 [C-2] and were calculated by plotting the 1/4T fluence values on Figure 2 of the Regulatory Guide and using the material-specific Cu wt. % values. The percent-loss lines were extended into the low fluence area of Regulatory Guide 1.99, Revision 2, Figure 2, i.e., below 1018 n/cm2, in order to determine the USE % decrease as needed.

(d)

Percentage USE decrease values are based on Position 2.2 of [C-2]. Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of [C-2]) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-5 WCAP-18746-NP October 2022 Revision 1 Figure C-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 1 1.0 10.0 100.0 1.00E+17 1.00E+18 1.00E+19 1.00E+20 Percentage Drop in USE Neutron Fluence, n/cm2 (E > 1 MeV)

Surveillance Material: Intermediate Shell Forging C Surveillance Material: Weld Heat # 1752

% Copper Base Metal Weld 0.35 0.30 0.30 0.25 0.25 0.20 0.20 0.15 0.15 0.10 0.10 0.05 Upper Limit Forging Line Weld Line Limiting Forging Percent USE Decrease 16% from Capsule N (Axial Orientation)

Limiting Weld Percent USE Decrease 9% from Capsule R Nozzle Shell Forging B 1/4T Fluence = 2.26x1019 n/cm2 Nozzle Shell to Intermediate Shell Circumferential Weld Seam - W2 1/4T Fluence = 2.43x1019 n/cm2 Lower Shell Forging D and Intermediate Shell to Lower Shell Circumferential Weld Seam - W3 1/4T Fluence = 3.70x1019 n/cm2 Intermediate Shell Forging C 1/4T Fluence = 3.77x1019 n/cm2

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-6 WCAP-18746-NP October 2022 Revision 1 Figure C-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 2 1.0 10.0 100.0 1.00E+17 1.00E+18 1.00E+19 1.00E+20 Percentage Drop in USE Neutron Fluence, n/cm2 (E > 1 MeV)

Surveillance Material: Lower Shell Forging D Surveillance Material: Weld Heat # 2721

% Copper Base Metal Weld 0.35 0.30 0.30 0.25 0.25 0.20 0.20 0.15 0.15 0.10 0.10 0.05 Upper Limit Forging Line Weld Line Limiting Forging Percent USE Decrease 13.8% from Capsule T (Tangential Orientation)

Limiting Weld Percent USE Decrease 11.7% from Capsule R Upper Shell Forging B 1/4T Fluence = 2.26x1019 n/cm2 Upper Shell to Intermediate Shell Circumferential Weld Seam - W2 1/4T Fluence = 2.42x1019 n/cm2 Lower Shell Forging D and Intermediate Shell to Lower Shell Circumferential Weld Seam - W3 1/4T Fluence = 3.73x1019 n/cm2 Intermediate Shell Forging C 1/4T Fluence = 3.79x1019 n/cm2

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-1 WCAP-18746-NP October 2022 Revision 1 APPENDIX D PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION D.1 PRESSURIZED THERMAL SHOCK Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the reactor pressure vessel (RPV) under the following conditions: severe overcooling of the inside surface of the vessel wall followed by high 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 [D-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 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 [D-2].

These accepted methods were used with the clad/base metal interface fluence values of Section 2 to calculate the following RTPTS values for the Prairie Island Units 1 and 2 RPV materials at 54 EFPY (EOLE).

The EOLE RTPTS calculations are summarized in Table D-1 and Table D-2.

PTS Conclusion The Prairie Island Units 1 and 2 limiting RTPTS value for base metal at 54 EFPY is 146.3°F (see Tables D-1 and D-2), which corresponds to Unit 1 Intermediate Shell Forging C when considering non-credible surveillance data (using Position 2.1). The Prairie Island Units 1 and 2 limiting RTPTS value for circumferentially oriented welds at 54 EFPY is 183.9°F, which corresponds to the Unit 1 Intermediate to Lower Shell Circumferential Weld Seam W3, Heat # 1752 considering non-credible surveillance data (using Position 2.1).

Therefore, all materials in the Prairie Island Units 1 and 2 reactor vessels are below the RTPTS screening criteria of 270F for base metal and/or longitudinal welds, and 300F for circumferentially oriented welds through EOLE (54 EFPY).

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-2 WCAP-18746-NP October 2022 Revision 1 Table D-1 RTPTS Calculations for the Prairie Island Unit 1 Reactor Vessel Materials at 54 EFPY Reactor Vessel Material CF(a)

(°F)

Surface Fluence(b)

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

Surface FF(b)

RTNDT(U)(c)

(°F)

RTNDT

(°F)

U

(°F)

(d)

(°F)

Margin

(°F)

RTPTS

(°F)

Nozzle Shell Forging B 51.0 3.37 1.318

-4 67.2 0

17.0 34.0 97.2 Intermediate Shell Forging C 44.0 5.63 1.425 14 62.7 0

17.0 34.0 110.7 Using Non-credible Prairie Island Unit 1 Surveillance Data 69.0 5.63 1.425 14 98.3 0

17.0 34.0 146.3 Lower Shell Forging D 44.0 5.53 1.422

-4 62.6 0

17.0 34.0 92.6 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 79.5 3.63 1.335 0

106.1 17 28.0 65.5 171.6 Intermediate to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 69.7 5.53 1.422

-13 99.1 0

28.0 56.0 142.1 Using Non-credible Prairie Island Unit 1 Surveillance Data 99.1 5.53 1.422

-13 140.9 0

28.0 56.0 183.9 Notes:

(a) Data is from Table 5-3.

(b) Data is from Table 7-1.

(c) Data is from Table 3-1.

(d) The credibility conclusion for the surveillance material is discussed in Section 4. The intermediate shell forging material surveillance data and the weld Heat # 1752 surveillance data were both determined to be non-credible. Per the guidance of 10 CFR 50.61 [D-1], the base metal = 17°F when surveillance data is non-credible or not used to determine the CF, and the weld metal = 28°F when surveillance data is not used to determine the CF and = 14°F when credible surveillance data is used to determine the CF. However, need not exceed 0.5*RTNDT per regulatory guidance in [D-1].

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-3 WCAP-18746-NP October 2022 Revision 1 Table D-2 RTPTS Calculations for the Prairie Island Unit 2 Reactor Vessel Materials at 54 EFPY Reactor Vessel Material CF(a)

(°F)

Surface Fluence(b)

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

Surface FF(b)

RTNDT(U)(c)

(°F)

RTNDT

(°F)

U

(°F)

(d)

(°F)

Margin

(°F)

RTPTS

(°F)

Upper Shell Forging B 44.0 3.37 1.318

-13 58.0 0

17.0 34.0 79.0 Intermediate Shell Forging C 44.0 5.66 1.426 14 62.7 0

17.0 34.0 110.7 Lower Shell Forging D 51.0 5.58 1.423

-4 72.6 0

17.0 34.0 102.6 Using Non-credible Prairie Island Unit 2 Surveillance Data 60.1 5.58 1.423

-4 85.5 0

17.0 34.0 115.5 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 69.7 3.61 1.334

-13 93.0 0

28.0 56.0 136.0 Using Non-credible Prairie Island Unit 1 Surveillance Data (Heat # 1752) 101.4 3.61 1.334

-13 135.2 0

28.0 56.0 178.2 Intermediate to Lower Shell Circumferential Weld - Seam W3 51.6 5.58 1.423

-31 73.4 0

28.0 56.0 98.4 Using Credible Prairie Island Unit 2 Surveillance Data (Heat # 2721) 80.8 5.58 1.423

-31 115.0 0

14.0 28.0 112.0 Notes:

(a) Data is from Table 5-4.

(b) Data is from Table 7-2.

(c) Data is from Table 3-2.

(d) The credibility conclusion for the surveillance material is discussed in Section 4. The lower shell forging material surveillance data was determined to be non-credible, while the weld Heat # 2721 surveillance data was determined to be credible. Additionally, as discussed in Section 4, the weld Heat # 1752 surveillance data was determined to be non-credible. Per the guidance of 10 CFR 50.61 [D-1], the base metal = 17°F when surveillance data is non-credible or not used to determine the CF, and the weld metal

= 28°F when surveillance data is not used to determine the CF and = 14°F when credible surveillance data is used to determine the CF. However, need not exceed 0.5*RTNDT per regulatory guidance in [D-1].

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-4 WCAP-18746-NP October 2022 Revision 1 D.2 EMERGENCY RESPONSE GUIDELINE LIMITS The Emergency Response Guideline (ERG) limits were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event [D-3]. 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 250F for a longitudinal flaw and 300F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250F for a longitudinal flaw or 300F 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 D.1 of this report. The material with the highest RTPTS defines the limiting material. Table D-3 and Table D-4 identify ERG category limits and the limiting material RTNDT values at 54 EFPY for Prairie Island Valley Units 1 and 2, respectively.

Table D-3 Evaluation of Prairie Island Unit 1 ERG Limit Category ERG Pressure-Temperature Limits [D-3]

Applicable RTNDT Value(a)

ERG P-T Limit Category RTNDT < 200F Category I 200F < RTNDT < 250F Category II 250F < RTNDT < 300F Category IIIb Limiting RTNDT Value(b)

Reactor Vessel Material RTNDT Value @ 54 EFPY Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752)

Using Non-credible Surveillance Data 183.9 Notes:

(a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.

(b) Value taken from Table D-1.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-5 WCAP-18746-NP October 2022 Revision 1 Table D-4 Evaluation of Prairie Island Unit 2 ERG Limit Category ERG Pressure-Temperature Limits [D-3]

Applicable RTNDT Value(a)

ERG P-T Limit Category RTNDT < 200F Category I 200F < RTNDT < 250F Category II 250F < RTNDT < 300F Category IIIb Limiting RTNDT Value(b)

Reactor Vessel Material RTNDT Value @ 54 EFPY Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 1752)

Using Non-credible Surveillance Data 178.2 Notes:

(a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.

(b) Value taken from Table D-2.

Per the ERG limit guidance document [D-3], 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.

Per Table D-3 and Table D-4, the limiting material for Prairie Island Units 1 and 2 have an RTNDT less than 200°F through 54 EFPY. Therefore, Prairie Island Units 1 and 2 remain in ERG Category I through EOLE (54 EFPY).

Conclusion of ERG P-T Limit Categorization As summarized above, Prairie Island Units 1 and 2 will remain in ERG Category I through EOLE (54 EFPY).

D.3 REFERENCES D-1 Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Federal Register, November 29, 2019.

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

D-3 Westinghouse Owners Group Document HF04BG, Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 3, March 31, 2014.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-1 WCAP-18746-NP October 2022 Revision 1 APPENDIX E VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS E.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for Capsules V, T, R, and P for Prairie Island Unit 2 are provided in this appendix. The sensor sets 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

[E-1]. One of the main purposes for providing 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 reported in Section 6.2.

E.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of Capsules V, T, R, and P for Prairie Island Unit 2 are presented.

The capsules designation, locations within the reactor, and time of withdrawal are as follows:

Capsule Azimuthal Location Withdrawal Time Irradiation Time (EFPY)

V 77º End of Cycle 1 1.39 T

67º End of Cycle 4 4.13 R

257º End of Cycle 9 8.80 P

247º End of Cycle 16 17.24 N(a) 237º End of Cycle 31 40.64 S

57º Standby Note:

(a) Capsule N has been removed but has not been analyzed yet. This report will be updated accordingly when the results of that evaluation is complete.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-2 WCAP-18746-NP October 2022 Revision 1 The passive neutron sensors included in these evaluations are summarized as follows:

Sensor Material Reaction of Interest Capsule V Capsule P Capsule R Capsule S Copper Cu-63 (n,) Co-60 X

X X

X Iron Fe-54 (n,p) Mn-54 X

X X

X Nickel Ni-58 (n,p) Co-58 X

X X

X Uranium-238 U-238 (n,f) Cs-137 X

X X

X Neptunium-237 Np-237 (n,f) Cs-137 X

X X

X Cobalt-Aluminum(a)

Co-59 (n,) Co-60 Note:

(a) For all the capsules withdrawn to date, none of the cobalt-aluminum wires have been recovered.

The design of the in-vessel surveillance capsules places the individual neutron sensors at several radial locations within the test specimen array. As a result of the various radial locations, gradient correction factors are applied to the measured reaction rates to index all of the neutron sensor measurements to a common geometric location (the center of the capsule) prior to use in the least-squares adjustment procedure. Pertinent physical and nuclear characteristics of the passive neutron sensors analyzed are listed in Table E-1.

The use of passive monitors does not yield a direct measure of the energy-dependent neutron exposure 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 exposure rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron exposure 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 The previously withdrawn in-vessel Capsules V, T, R, and P for Prairie Island Unit 2 were re-evaluated using the current calculational model.

The operating history of the reactor over the irradiation periods was based on the monthly power generation of Prairie Island Unit 2 from initial reactor criticality through the end of the dosimetry evaluation period.

For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history for Cycle 1 through Cycle 16 is in [E-2].

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-3 WCAP-18746-NP October 2022 Revision 1 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).

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 variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The Ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 2.2, accounts for the change in sensor reaction rates caused by variations in exposure rate level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, the additional Cj term should be employed. The impact of changing exposure rate levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low-leakage to low-leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another. The fuel-cycle-specific neutron exposure rate values are

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-4 WCAP-18746-NP October 2022 Revision 1 used to compute cycle-dependent values for Cj values at the radial and azimuthal center of the respective capsules at core midplane.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the U-238 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 U-238 and Np-237 sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the surveillance capsule irradiations. The correction factors corresponding to the Prairie Island Unit 2 fission sensor reaction rates are summarized as follows:

Correction Capsule V Capsule T Capsule R Capsule P U-235 Impurity/Pu Build-in 0.861 0.842 0.7389 0.733 U-238 (,f) 0.955 0.960 0.955 0.960 Net U-238 Correction 0.822 0.808 0.706 0.704 Np-237 (,f) 0.985 0.986 0.985 0.986 The correction factors were applied in a multiplicative fashion to the decay-corrected cadmium-covered fission sensor reaction rates.

Results of the sensor reaction rate determinations for the in-vessel Capsules V, T, R, and P are given in Table E-2 through Table E-5, where the measured specific activities, decay-corrected saturated specific activities, and computed reaction rates for each sensor are listed.

E.1.2 Least-Squares Evaluation of Sensor Sets Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best-estimates for key exposure parameters such as fluence rate (E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares method, as applied to 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,

)

)(

(

R g

ig i

g

g ig R

i

=

relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross-sections, 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.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-5 WCAP-18746-NP October 2022 Revision 1 For the least-squares evaluation of the Prairie Island Unit 2 dosimetry, the FERRET code [E-3] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine the best-estimate values of exposure parameters (fluence rate (E > 1.0 MeV) and dpa) and their 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 rates and associated uncertainty for each sensor contained in the multiple sensor set.

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

For Prairie Island Unit 2, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 2.2. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section E.1.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [E-4].

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 E944, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance [E-5].

The following provides a summary of the uncertainties associated with the least-squares evaluation of the Prairie Island Unit 2 surveillance capsule sensor sets.

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

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

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-6 WCAP-18746-NP October 2022 Revision 1 Reaction Uncertainty 63Cu (n,) 60Co 5%

54Fe (n,p) 54Mn 5%

58Ni (n,p) 58Co 5%

59Co (n,) 60Co 5%

238U (n,f) FP 10%

237Np (n,f) FP 10%

These uncertainties are given at the 1 level.

Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least-squares evaluations were taken from the SNLRML library.

This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least-squares adjustment applications. These cross-sections were compiled from recent cross-section evaluations, and they have been tested for 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 Prairie Island Unit 2 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package:

Reaction Uncertainty Cu-63 (n,) Co-60 4.08-4.16%

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

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

Co-59 (n,) Co-60 0.79-3.59%

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

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

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

Calculated Neutron Spectrum The neutron spectra inputs 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).

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-7 WCAP-18746-NP October 2022 Revision 1 Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

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

gg' g'

g 2

n gg' P

R R

R M

+

=

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:

e +

]

[1

=

P

-H g

g g

g

where:

2 2

2

)

g' (g

H

=

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 Prairie Island Unit 2 calculated spectra was as follows:

Exposure Rate Normalization Uncertainty (Rn) 15%

Exposure Rate Group Uncertainties (Rg, Rg')

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlation ()

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

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-8 WCAP-18746-NP October 2022 Revision 1 E.1.3 Comparisons of Measurements and Calculations Results of the least-squares evaluations are provided in Table E-6 through Table E-9. In these tables, measured, calculated, and best-estimate values for sensor reaction rates are given. Also provided in these tabulations are ratios of the measured reaction rates to both the calculated and least-squares adjusted reaction rates. These ratios of measured-to-calculated (M/C) and measured-to-best estimate (M/BE) illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. Additionally, comparisons of the calculated and best-estimate values of neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the best-estimate-to-calculated (BE/C) ratios observed for each of the capsules.

The data comparisons provided in Table E-6 through Table E-9 show that the adjustments to the calculated spectra are relatively small and within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least-squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 2.4, the calculational uncertainty is specified as 13% at the 1 level.

Further comparisons of the measurement results with calculations are given in Table E-10 and Table E-11.

In Table E-10, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table E-11, calculations of fast neutron exposure rates in terms of fast neutron (E > 1.0 MeV) fluence rate and dpa/s are compared with the best-estimate results obtained from the least-squares evaluation of the capsule dosimetry results. These comparisons yield consistent and similar results with all measurement-to-calculation comparisons falling within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of the measured and calculated sensor reaction rates, for the individual threshold sensors considered in the least-squares analysis, the M/C comparisons of the fast neutron threshold reactions range from 0.93 to 1.11. The overall average M/C ratio is 0.99 with an associated standard deviation of 9.0%.

In the case of the comparison of the best-estimate and calculated fast neutron exposure parameters, the BE/C comparisons are 0.98 and 0.99 for fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.2 are valid for use in the assessment of the condition of the materials comprising the beltline region of the Prairie Island Unit 2 reactor pressure vessel.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-9 WCAP-18746-NP October 2022 Revision 1 Table E-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors Reaction of Interest Atomic Weight (g/g-atom)

Target Atom Fraction Product Half-life (days)

Fission Yield

(%)

90%

Response

Range(a)

(MeV)

Cu-63 (n,) Co-60 63.546 0.6917 1925.28 4.53-11.0 Fe-54 (n,p) Mn-54 55.845 0.05845 312.13 2.27-7.54 Ni-58 (n,p) Co-58 58.693 0.68077 70.86 1.98-7.51 Co-59 (n,) Co-60 58.933 0.0015 1925.28 Non-threshold U-238 (n,f) Cs-137 238.051 1.00 10975.76 6.02 1.44-6.69 Np-237 (n,f) Cs-137 237.048 1.00 10975.76 6.27 0.68-5.61 Note:

(a) Energies between which 90% of activity is produced (U-235 fission spectrum) [E-6].

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-10 WCAP-18746-NP October 2022 Revision 1 Table E-2 Measured Sensor Activities and Reaction Rates for Surveillance Capsule V Sensor Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Reaction Rate (rps/atom)

Average Reaction Rate (rps/atom)

Corrected Average Reaction Rate (rps/atom)

Cu Top Middle 5.900E+04 4.320E+05 6.591E-17 6.903E-17 6.903E-17 Cu Bottom Middle 6.460E+04 4.730E+05 7.216E-17 Iron Top 2.370E+06 5.045E+06 8.004E-15 7.997E-15 7.997E-15 Iron Top Middle 2.210E+06 4.704E+06 7.464E-15 Iron Middle 2.340E+06 4.981E+06 7.903E-15 Iron Bottom Middle 2.430E+06 5.173E+06 8.207E-15 Iron Bottom 2.490E+06 5.300E+06 8.409E-15 Nickel Middle 1.397E+07 7.543E+07 1.080E-14 1.080E-14 1.080E-14 U-238 (Cd)

Middle 2.620E+05 8.394E+06 5.512E-14 5.512E-14 4.532E-14 Np-237 (Cd)

Middle 2.250E+06 7.209E+07 4.526E-13 4.526E-13 4.459E-13 Bare Co Top 2.150E+07 1.277E+08 8.334E-12 8.082E-12 8.082E-12 Bare Co Bottom 2.020E+07 1.200E+08 7.830E-12 Note:

(a) Measured activity is decay corrected to 3/14/1977.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-11 WCAP-18746-NP October 2022 Revision 1 Table E-3 Measured Sensor Activities and Reaction Rates for Surveillance Capsule T Sensor Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Reaction Rate (rps/atom)

Average Reaction Rate (rps/atom)

Corrected Average Reaction Rate (rps/atom)

Cu Top Middle 1.040E+05 3.315E+05 5.058E-17 5.350E-17 5.350E-17 Cu Bottom Middle 1.160E+05 3.698E+05 5.641E-17 Iron Top 1.650E+06 3.647E+06 5.786E-15 5.681E-15 5.681E-15 Iron Top Middle 1.470E+06 3.249E+06 5.155E-15 Iron Middle 1.650E+06 3.647E+06 5.786E-15 Iron Bottom Middle 1.610E+06 3.559E+06 5.646E-15 Iron Bottom 1.720E+06 3.802E+06 6.032E-15 Nickel Middle 1.790E+06 5.732E+07 8.206E-15 8.206E-15 8.206E-15 U-238 (Cd)

Middle 4.840E+05 5.471E+06 3.592E-14 3.592E-14 2.901E-14 Np-237 (Cd)

Middle 4.090E+06 4.623E+07 2.902E-13 2.902E-13 2.860E-13 Bare Co Top 2.650E+07 6.682E+07 4.360E-12 4.319E-12 4.319E-12 Bare Co Bottom 2.600E+07 6.556E+07 4.277E-12 Note:

(a) Measured activity is decay corrected to 12/2/1980.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-12 WCAP-18746-NP October 2022 Revision 1 Table E-4 Measured Sensor Activities and Reaction Rates for Surveillance Capsule R Sensor Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Reaction Rate (rps/atom)

Average Reaction Rate (rps/atom)

Corrected Average Reaction Rate (rps/atom)

Cu Top Middle 2.380E+05 4.768E+05 7.274E-17 7.549E-17 7.549E-17 Cu Bottom Middle 2.560E+05 5.129E+05 7.824E-17 Iron Top 2.642E+06 5.404E+06 8.574E-15 8.567E-15 8.567E-15 Iron Top Middle 2.431E+06 4.973E+06 7.889E-15 Iron Middle 2.602E+06 5.322E+06 8.444E-15 Iron Bottom Middle 2.525E+06 5.165E+06 8.194E-15 Iron 2.851E+06 5.832E+06 9.252E-15 Iron Bottom 2.789E+06 5.705E+06 9.051E-15 Nickel Middle 3.320E+06 8.237E+07 1.179E-14 1.179E-14 1.179E-14 U-238 (Cd)

Middle 2.165E+06 1.222E+07 8.025E-14 8.025E-14 5.661E-14 Np-237 (Cd)

Middle 1.282E+07 7.237E+07 4.543E-13 4.543E-13 4.475E-13 Bare Co Top 6.924E+07 1.126E+08 7.343E-12 7.703E-12 7.703E-12 Bare Co Bottom 7.604E+07 1.236E+08 8.064E-12 Note:

(a) Measured activity is decay corrected to 7/22/1986.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-13 WCAP-18746-NP October 2022 Revision 1 Table E-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule P Sensor Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Reaction Rate (rps/atom)

Average Reaction Rate (rps/atom)

Corrected Average Reaction Rate (rps/atom)

Cu Top Middle 1.980E+05 3.226E+05 4.921E-17 5.120E-17 5.120E-17 Cu Bottom Middle 2.140E+05 3.487E+05 5.319E-17 Iron Top 1.930E+06 3.312E+06 5.254E-15 5.053E-15 5.053E-15 Iron Top Middle 1.650E+06 2.831E+06 4.492E-15 Iron Middle 1.850E+06 3.174E+06 5.036E-15 Iron Bottom Middle 1.780E+06 3.054E+06 4.846E-15 Iron Bottom 2.070E+06 3.552E+06 5.635E-15 Nickel Middle 7.830E+06 4.984E+07 7.135E-15 7.135E-15 7.135E-15 U-238 (Cd)

Middle 1.690E+06 5.421E+06 3.560E-14 3.560E-14 2.507E-14 Np-237 (Cd)

Middle 1.160E+07 3.721E+07 2.336E-13 2.336E-13 2.302E-13 Bare Co Top 3.930E+07 5.065E+07 3.304E-12 3.346E-12 3.346E-12 Bare Co Bottom 4.030E+07 5.193E+07 3.388E-12 Note:

(a) Measured activity is decay corrected to 10/20/1995.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-14 WCAP-18746-NP October 2022 Revision 1 Table E-6 Least-Squares Evaluation of Dosimetry in Surveillance Capsule V (13° Position, Core Midplane, Irradiated During Cycle 1) 2/DOF = 0.199 Reaction Rate (rps/atom)

M/C M/BE BE/C Reaction Measured (M)

Calculated (C)

Best-Estimate (BE) 63Cu (n,) 60Co 6.90E-17 7.02E-17 6.78E-17 0.98 1.02 0.97 54Fe (n,p) 54Mn 8.00E-15 8.47E-15 8.06E-15 0.94 0.99 0.95 58Ni (n,p) 58Co 1.08E-14 1.18E-14 1.12E-14 0.91 0.97 0.94 238U(Cd) (n,f) 137Cs 4.53E-14 4.50E-14 4.36E-14 1.01 1.04 0.97 237Np(Cd) (n,f) 137Cs 4.46E-13 3.95E-13 4.18E-13 1.13 1.06 1.06 59Co (n,) 60Co 8.08E-12 1.01E-11 8.13E-12 0.80 0.99 0.80 Average of Fast Energy Threshold Reactions 0.99 1.02 0.98 Percent Standard Deviation 8.6 3.6 4.9 Integral Quantity Calculated (C)

% Unc.

Best-Estimate (BE)

% Unc.

BE/C Neutron Fluence Rate (E > 1.0 MeV)

(n/cm2-s) 1.37E+11 13 1.34E+11 6

0.98 Displacement Rate (dpa/s) 2.46E-10 13 2.44E-10 7

0.99

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-15 WCAP-18746-NP October 2022 Revision 1 Table E-7 Least-Squares Evaluation of Dosimetry in Surveillance Capsule T (23° Position, Core Midplane, Irradiated During Cycles 1 through 4) 2/DOF = 0.176 Reaction Rate (rps/atom)

M/C M/BE BE/C Reaction Measured (M)

Calculated (C)

Best-Estimate (BE) 63Cu (n,) 60Co 5.35E-17 5.62E-17 5.34E-17 0.95 1.00 0.95 54Fe (n,p) 54Mn 5.68E-15 6.04E-15 5.83E-15 0.94 0.97 0.96 58Ni (n,p) 58Co 8.21E-15 8.30E-15 8.14E-15 0.99 1.01 0.98 238U(Cd) (n,f) 137Cs 2.90E-14 2.93E-14 2.93E-14 0.99 0.99 1.00 237Np(Cd) (n,f) 137Cs 2.86E-13 2.34E-13 2.63E-13 1.22 1.09 1.12 59Co (n,) 60Co 4.32E-12 5.16E-12 4.34E-12 0.84 1.00 0.84 Average of Fast Energy Threshold Reactions 1.02 1.01 1.00 Percent Standard Deviation 11.3 4.5 6.9 Integral Quantity Calculated (C)

% Unc.

Best-Estimate (BE)

% Unc.

BE/C Neutron Fluence Rate (E > 1.0 MeV)

(n/cm2-s) 8.49E+10 13 8.68E+10 6

1.02 Displacement Rate (dpa/s) 1.46E-10 13 1.51E-10 7

1.03

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-16 WCAP-18746-NP October 2022 Revision 1 Table E-8 Least-Squares Evaluation of Dosimetry in Surveillance Capsule R (13° Position, Core Midplane, Irradiated During Cycles 1 through 9) 2/DOF = 0.684 Reaction Rate (rps/atom)

M/C M/BE BE/C Reaction Measured (M)

Calculated (C)

Best-Estimate (BE) 63Cu (n,) 60Co 7.55E-17 7.62E-17 7.39E-17 0.99 1.02 0.97 54Fe (n,p) 54Mn 8.57E-15 9.19E-15 8.80E-15 0.93 0.97 0.96 58Ni (n,p) 58Co 1.18E-14 1.28E-14 1.22E-14 0.92 0.96 0.95 238U(Cd) (n,f) 137Cs 5.66E-14 4.89E-14 4.82E-14 1.16 1.18 0.99 237Np(Cd) (n,f) 137Cs 4.47E-13 4.30E-13 4.39E-13 1.04 1.02 1.02 59Co (n,) 60Co 7.70E-12 1.10E-11 7.78E-12 0.70 0.99 0.71 Average of Fast Energy Threshold Reactions 1.01 1.03 0.98 Percent Standard Deviation 9.7 8.6 2.8 Integral Quantity Calculated (C)

% Unc.

Best-Estimate (BE)

% Unc.

BE/C Neutron Fluence Rate (E > 1.0 MeV)

(n/cm2-s) 1.48E+11 13 1.48E+11 6

1.00 Displacement Rate (dpa/s) 2.67E-10 13 2.67E-10 7

1.00

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-17 WCAP-18746-NP October 2022 Revision 1 Table E-9 Least-Squares Evaluation of Dosimetry in Surveillance Capsule P (23° Position, Core Midplane, Irradiated During Cycles 1 through 16) 2/DOF = 0.139 Reaction Rate (rps/atom)

M/C M/BE BE/C Reaction Measured (M)

Calculated (C)

Best-Estimate (BE) 63Cu (n,) 60Co 5.12E-17 5.36E-17 5.03E-17 0.96 1.02 0.94 54Fe (n,p) 54Mn 5.05E-15 5.68E-15 5.18E-15 0.89 0.97 0.91 58Ni (n,p) 58Co 7.13E-15 7.79E-15 7.16E-15 0.92 1.00 0.92 238U(Cd) (n,f) 137Cs 2.51E-14 2.73E-14 2.53E-14 0.92 0.99 0.93 237Np(Cd) (n,f) 137Cs 2.30E-13 2.17E-13 2.17E-13 1.06 1.06 1.00 59Co (n,) 60Co 3.35E-12 4.73E-12 3.38E-12 0.71 0.99 0.71 Average of Fast Energy Threshold Reactions 0.95 1.01 0.94 Percent Standard Deviation 7.0 3.4 3.8 Integral Quantity Calculated (C)

% Unc.

Best-Estimate (BE)

% Unc.

BE/C Neutron Fluence Rate (E > 1.0 MeV)

(n/cm2-s) 7.88E+10 13 7.37E+10 6

0.93 Displacement Rate (dpa/s) 1.35E-10 13 1.28E-10 7

0.95

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-18 WCAP-18746-NP October 2022 Revision 1 Table E-10 Measured-to-Calculated (M/C) Reaction Rates - In-Vessel Capsules Reaction Capsule Average

% Std. Dev.

V T

R P

63Cu (n,) 60Co 0.98 0.95 0.99 0.96 0.97 1.9 54Fe (n,p) 54Mn 0.94 0.94 0.93 0.89 0.93 2.6 58Ni (n,p) 58Co 0.91 0.99 0.92 0.92 0.94 4.0 238U(Cd) (n,f) 137Cs 1.01 0.99 1.16 0.92 1.02 9.9 237Np(Cd) (n,f) 137Cs 1.13 1.22 1.04 1.06 1.11 7.3 Average of M/C Results 0.99 9.0 Table E-11 Best-Estimate-to-Calculated (BE/C) Exposure Rates - In-Vessel Capsules Capsule Neutron Fluence (E > 1.0 MeV)

Rate BE/C Iron Atom Displacement Rate BE/C V

0.98 0.99 T

1.02 1.03 R

1.00 1.00 P

0.93 0.95 Average 0.98 0.99

% Std. Dev.

3.9 3.3

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-19 WCAP-18746-NP October 2022 Revision 1 E.2 REFERENCES E-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [ADAMS Accession Number ML010890301]

E-2 Westinghouse Report WCAP-14613, Revision 2, Analysis of Capsule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, February 1998.

E-3 A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

E-4 RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.

E-5 ASTM Standard E944-19, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, 2019.

E-6 ASTM Standard E844-18, Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, 2018.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-1 WCAP-18746-NP October 2022 Revision 1 APPENDIX F CREDIBILITY EVALUATION OF THE PRAIRIE ISLAND UNIT 2 SURVEILLANCE PROGRAM F.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [F-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes 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 Position C.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 Prairie Island Unit 2 reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Prairie Island Unit 2 reactor vessel surveillance data and determine if that surveillance data is credible.

F.2 EVALUATION 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 Part 50, "Fracture Toughness Requirements," [F-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."

In addition, the Upper Shell Forging B and the Intermediate Shell Forging to Upper Shell Forging Weld Seam W2 will be considered to be part of the beltline region. Hence, the Prairie Island Unit 2 reactor vessel beltline region consists of the following materials:

1. Upper Shell Forging B (Heat # 22231/39088).

2. Intermediate Shell Forging C (Heat # 22829).

3. Lower Shell Forging D (Heat # 22642).

4. Upper Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2 (Weld Wire Type UM40/Heat # 1752, Flux Type UM 89/Lot # 1263).

5. Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3 (Weld Wire Type UM40/Heat # 2721, Flux Type UM 89/Lot # 1263).

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-2 WCAP-18746-NP October 2022 Revision 1 Per WCAP-8193, Revision 0 [F-3], the Prairie Island Unit 2 surveillance program was based on ASTM E185-70 [F-4], Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Per Section 3.1.2 of ASTM E185-70, A minimum test program shall consist of specimens taken from the following locations: (1) base metal of one heat, incorporated in the highest flux location of the reactor vessel, that has the highest initial ductile-brittle transition temperature, (2) weld metal fully representative of fabrication practice used for the welds in the highest flux location of the reactor vessel (weld wire or rod, and flux must come from one of the heats used in the highest flux region of the reactor vessel), and (3) the heat-affected zone of the weldments noted above.

It should be noted here that the Upper Shell Forging B and Weld Seam W2 were not considered when the surveillance program was developed. Therefore, at the time the Prairie Island Unit 2 surveillance capsule program was developed, Lower Shell Forging D was judged to be most limiting. This was based on the fact that both the Intermediate Shell Forging C and Lower Shell Forging D initial RTNDT values were within 2F of each other and Lower Shell Forging D had a lower USE value. As for the weld, the Prairie Island Unit 2 vessel has only one weld in the highest flux region (Weld Seam W3, Weld Control # PS-011, Type UM40, Heat # 2721, Flux Type UM89, Flux Lot No. 1263). The same weld was used in the surveillance program.

Therefore, the materials selected for use in the Prairie Island Unit 2 surveillance program were those judged to be most likely limiting with regard to radiation embrittlement according to the accepted methodology at the time the surveillance program was developed.

Based on the discussion, Criterion 1 is met for the Prairie Island Unit 2 surveillance program.

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.

Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions are presented in Section 5 and Appendix C of the latest surveillance capsule report, WCAP-14613 [F-5].

Based on engineering judgment, the scatter in the data presented in these plots, as documented in Appendix A of this calculation note, is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Prairie Island Unit 2 surveillance materials unambiguously.

Hence, Criterion 2 is met for the Prairie Island Unit 2 surveillance program.

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 [F-6].

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 forging.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 [F-7]. At this meeting the NRC presented five cases. Of the five cases, Case 1

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-3 WCAP-18746-NP October 2022 Revision 1 (Surveillance data available from plant but no other source) most closely represents the situation listed above for Prairie Island Unit 2 surveillance weld metal and forging materials. It is noted that the Unit 2 Upper Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2 utilizes surveillance data for Heat # 1752 from the Prairie Island Unit 1 surveillance program, consistent with Case 5 (Surveillance data from other sources only); however, creditability of this data was evaluated in WCAP-18660-NP [F-8].

Table F-1 contains the calculation of chemistry factors for the Prairie Island Unit 2 reactor vessel beltline materials contained in the surveillance program.

Table F-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Prairie Island Unit 2 Surveillance Data Material Capsule Capsule Fluence(a)

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

FF(b)

RTNDT(c)

(°F)

FF*RTNDT

(°F)

FF(b)

Lower Shell Forging D (Axial)

V 0.598 0.856 35.28 30.20 0.733 T

1.10 1.027 29.93 30.73 1.054 R

4.11 1.362 84.73 115.41 1.855 P

4.27 1.370 103.87 142.32 1.877 Lower Shell Forging D (Tangential)

V 0.598 0.856 32.89 28.15 0.733 T

1.10 1.027 55.69 57.17 1.054 R

4.11 1.362 90.02 122.61 1.855 P

4.27 1.370 99.91 136.89 1.877 SUM:

663.5 11.039 CF LS D = (FF

RTNDT) ÷ (FF2) = (663.5) ÷ (11.039) = 60.1°F Surveillance Weld (Heat #2721)

V 0.598 0.856 70.07 59.98 0.733 T

1.10 1.027 57.73 59.27 1.054 R

4.11 1.362 100.31 136.63 1.855 P

4.27 1.370 96.24 131.87 1.877 SUM:

387.7 5.519 CF Surv. Weld = (FF

RTNDT) ÷ (FF2) = (387.7) ÷ (5.519) = 70.3°F Notes:

(a) Fluence taken from Table 2-10.

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

(c) These measured RTNDT values do not include the adjustment ratio procedure of Reg. Guide 1.99 Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values and based on the copper and nickel content the ratio would be 1. In addition, the only surveillance data available is from the Prairie Island Unit 2 reactor vessel; therefore, no temperature adjustment is required.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-4 WCAP-18746-NP October 2022 Revision 1 The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table F-2.

Table F-2 Best-Fit Evaluation for Prairie Island Unit 2 Surveillance Materials Material Capsule CF(a)

(Slopebest-fit)

(°F)

Capsule Fluence(b)

(x 1019 n/cm2)

FF(c)

Measured RTNDT(d)

(°F)

Predicted RTNDT(e)

(°F)

Scatter RTNDT(f)

(°F)

<17°F (Base Metal)

<28°F (Weld)

Lower Shell Forging D (Axial)

V 60.1 0.598 0.856 35.28 51.5 16.2 Yes T

1.10 1.027 29.93 61.7 31.8 No R

4.11 1.362 84.73 81.9 2.9 Yes P

4.27 1.370 103.87 82.4 21.5 No Lower Shell Forging D (Tangential)

V 0.598 0.856 32.89 51.5 18.6 No T

1.10 1.027 55.69 61.7 6.0 Yes R

4.11 1.362 90.02 81.9 8.2 Yes P

4.27 1.370 99.91 82.4 17.6 No Surveillance Weld Metal (Heat # 2721)

V 70.3 0.598 0.856 70.1 60.1 9.9 Yes T

1.10 1.027 57.7 72.1 14.4 Yes R

4.11 1.362 100.3 95.7 4.6 Yes P

4.27 1.370 96.2 96.3 0.0 Yes Notes:

(a) CF calculated in Table F-1.

(b) Fluence taken from Table 2-10.

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

(d) Measured RTNDT taken from Table F-1.

(e) Predicted RTNDT = CF x FF.

(f) Scatter RTNDT = Absolute Value [Predicted RTNDT - Adjusted RTNDT].

The scatter of RTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 17°F for base metal and 28°F for weld metal. From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table F-2 indicates that four of the eight surveillance data points fall inside the +/- 1 of 17°F scatter band for surveillance base metals, which is 50% of the data (4/8 x 100). In addition, Table F-2 indicates that zero of the four surveillance data points fall outside the +/- 1 of 28°F scatter band for surveillance weld materials.

Therefore, the Lower Shell Forging D is deemed not-credible, while the Surveillance weld is deemed credible. Although Lower Shell Forging D did not meet Criterion 3, both materials may still be used in determining the upper-shelf energy decrease in accordance with Regulatory Guide 1.99, Revision 2, Position 2.2.

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 Prairie Island Unit 2 capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-5 WCAP-18746-NP October 2022 Revision 1 the thermal shield. 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 and will not differ by more than 25F.

The weld metal used in Weld Seam W2 is contained in the Unit 1 surveillance program. Both Prairie Island Unit 1 and Unit 2 operate at the same temperature. Hence, the Unit 1 surveillance program weld metal was irradiated at a temperature within +/- 25F of weld seam W2 in Unit 2. Therefore, the Unit 1 surveillance weld will be used to project the fracture toughness properties of the Unit 2 Weld Seam W2.

Hence, Criterion 4 is met for the Prairie Island Unit 2 surveillance program.

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 Prairie Island Unit 2 surveillance program does contain correlation monitor material, which was supplied by Oak Ridge National Laboratory from plate material used in the AEC-Sponsored Heavy Section Steel Technology (HSST) Program. This material was obtained from a 12-inch-thick A533, Grade B, Class 1 plate (HSST Plate 02), which was provided to Subcommittee II of ASTM Committee E10 on Radioisotopes and Radiation Effects to serve as correlation monitor material in reactor vessel surveillance programs. The plate was produced by the Lukens Steel Company and heat treated by Combustion Engineering, Inc.

This criterion was evaluated and met previously in WCAP-14613 [F-5]. However, this criterion will be re-evaluated using the updated surveillance capsule fluence values. NUREG/CR-6413, ORNL/TM-13133 [F-9], contains a plot of Residual vs. Fast Fluence for the HSST01 and HSST02 correlation monitor material (Figure 11 of the report). The figure shows a 2 uncertainty of 50°F. The data used for this plot is contained in Tables 13, 14, and 15 (in the NUREG Report). However, the data in the NUREG report does not consider the recalculated fluence values documented herein. Thus, Table F-3 below presents an updated calculation of Residual vs. Fast Fluence for Prairie Island Unit 2.

Table F-3 Calculation of Residual vs. Fast Fluence for Prairie Island Unit 2 Capsule Capsule Fluence (x 1019 n/cm2, E > 1.0 MeV)

FF Measured Shift(a)

(°F)

RG 1.99 Shift(b)

(°F)

Residual(c)

V 0.598 0.856 123.58 109.6 14.0 T

1.10 1.027 158.14 131.4 26.7 R

4.11 1.362 183.19 174.3 8.8 P

4.27 1.370 196.3 175.4 20.9 Notes:

(a) Measured T30 values for the correlation monitor material were taken from Table B-17 of WCAP-14613 [F-5] for HSST02 material.

(b) Per NUREG/CR-6413, ORNL/TM-13133 [F-9], the Cu and Ni values for the correlation monitor material (HSST Plate

02) are 0.17 and 0.64, respectively. This equates to a chemistry factor value of 128F based on Regulatory Guide 1.99, Revision 2, Position 1.1. The calculated shift is thus equal to CF

FF.

(c) Residual = Measured Shift - RG 1.99 Shift.

Table F-3 shows a 2 uncertainty of less than 50F, which is the allowable scatter in NUREG/CR-6413, ORNL/TM-13133.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-6 WCAP-18746-NP October 2022 Revision 1 Hence, Criterion 5 is met for the Prairie Island Unit 2 surveillance program.

F.3 CONCLUSION Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgment, the Prairie Island Unit 2 surveillance forging material is deemed non-credible and the Prairie Island Unit 2 surveillance weld material is deemed credible.

- - This record was final approved on 10/17/2022, 8:03:18 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-7 WCAP-18746-NP October 2022 Revision 1 F.4 REFERENCES F-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

[ADAMS Accession Number ML003740284]

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

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

F-3 Westinghouse Report WCAP-8193, Revision 0, Northern States Power Co. Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program, September 1973.

F-4 ASTM E185-70, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, American Society of Testing and Materials, Philadelphia, PA, 1970.

F-5 Westinghouse Report WCAP-14613, Revision 2, Analysis of Capsule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, February 1998.

F-6 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society of Testing and Materials, Philadelphia, PA, 1982.

F-7 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]

F-8 Westinghouse Report WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, November 2021.

F-9 NUREG/CR-6413; ORNL/TM-13133, Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials, J. A. Wang, Oak Ridge National Laboratory, Oak Ridge, TN, April 1996. [ADAMS Accession Number ML20112B397].

ENCLOSURE 3 Westinghouse Analysis NSPM-LTP-TR-AA-000001-NP, Revision 1 Prairie Island Units 1 and 2 Low Temperature Overpressure Protection System (LTOPS) Analysis

[NON-PROPRIETARY]

28 pages follow

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 1 of 28 NSPM-LTP-TR-AA-000001-NP Rev. 1 Prairie Island Units 1 and 2 Low Temperature Overpressure Protection System (LTOPS) Analysis Andrew D. Sippel Thomas G. Joseph Verified By:

Bryan D. Jaskiewicz August 2022

- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 2 of 28 1.0 Introduction The Low Temperature Overpressure Protection System (LTOPS), also known as the Overpressure Protection System (OPPS), provides Reactor Coolant System (RCS) pressure relief capability during relatively low temperature operation to minimize the potential for challenging reactor vessel integrity limits (i.e., 10 CFR 50, Appendix G limits). At Prairie Island, in accordance with Technical Specification (TS) Limiting Condition for Operation (LCOs) 3.4.12 and 3.4.13, the pressurizer Power Operated Relief Valves (PORVs),

with reduced lift settings, provide a method of LTOP for the potential overpressure transients that may occur at low temperature RCS conditions. The LTOPS PORV setpoints are selected in accordance with the NRC approved methodology in Reference 10 such that the peak pressure during the design basis Mass Injection (MI) and Heat Injection (HI) transients will not exceed the Isothermal Appendix G P-T Limit Curve. The design basis MI and HI transients analyzed herein for Prairie Island Units 1 and 2 are as follows:

MI: This transient assumes a system failure, whereby the charging line flow control valve fails open and, simultaneously, the letdown line flow control valve fails closed. The limiting mass input is based on:

o Cold Leg Temperature 200°F (analytical limit)

The limiting MI event is based on all three charging pumps injecting.

o 200°F < Cold Leg Temperature LTOPS Enable Temperature The limiting MI event results from one SI pump plus three charging pumps injecting.

HI (without steam bubble): This transient assumes the inadvertent start-up of an RCP while the primary system is in a water solid condition and the Steam Generator (SG) secondary sides are at a temperature of 50°F higher than the remainder of the primary RCS.

HI (with steam bubble): This transient assumes the inadvertent start-up of an RCP while the pressurizer level is 90% and the SG secondary sides are at a temperature of 150°F higher than the remainder of the primary RCS.

The current P-T limits are valid to 54 Effective Full Power Years (EFPY) for Units 1 and 2 and the most recent LTOPS evaluation was performed in Reference 2. In accordance with References 13 and 14, Westinghouse has developed new P-T limit curves in Reference 15 at 54 EFPY that bound both Prairie Island Units 1 and 2. Westinghouse has also performed an LTOPS analysis to determine maximum allowable pressurizer PORV lift settings necessary to protect these new P-T limits. This report summarizes the results of this LTOPS analysis.

1.1 Limits of Applicability The results of this report are applicable to Prairie Island Units 1 & 2 operation with the P-T limit curves to the End of License Extension (EOLE) term of 54 EFPY defined in References 3 and 15. The key analysis inputs for this analysis are defined in Section 2.0.

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 3 of 28 2.0 Input Parameters and Assumptions Key input parameters for the analysis were requested in Reference 4 and the Xcel response was provided in Reference 5. The key input parameters and analysis assumptions are summarized as follows.

2.1 Key Inputs Design Basis MI Transient Due to restrictions on pump operability, the MI transient analysis for Prairie Island was performed for two distinct temperature ranges.

a.

Cold Leg Temperature 200°F (analytical limit)

Below 218 °F, which includes an 18 °F temperature uncertainty, References 6 and 7 require both Safety Injection (SI) pumps to be incapable of injecting into the RCS. Therefore, the limiting MI event is all three charging pumps injecting into a water solid RCS with letdown isolated.

b.

200°F < Cold Leg Temperature LTOPS Enable Temperature(1)

Reference 6 allows a maximum of one SI pump capable of injecting into the RCS in this temperature range. Therefore, the limiting MI event results from one SI pump plus three charging pumps injecting into a water solid RCS with letdown isolated.

Note 1: The current LTOPS enable temperature is 310 °F and above this temperature LTOPS is not needed for RCS overpressurization protection. This temperature value of 310 °F was used in Reference 8; however, it has been requested that Westinghouse evaluate a reduction of the enable temperature to 290 °F as part of this project. See Section 5.4 for the enable temperature evaluation.

MI Flow Rate Per Reference 5, MI pump flow rates are as follows:

three charging pumps

= [

]a,c gpm (maximum of [

]a,c gpm per pump)

SI pump

= [

]a,c gpm (maximum runout flow of any SI pump)

The PORV setpoint overshoots and undershoots were analyzed for MI flow rates of [

]a,c gpm (corresponding to three charging pumps injecting) and [

]a,c gpm (corresponding to three charging pumps injecting and one SI pump injecting) with PORV setpoints ranging from [

]a,c psig. The results of all MI parametric analyses in Reference 9 are summarized in Table A-1.

Design Basis HI Transient The HI transient is defined as the startup of one RCP with the SG secondary side a maximum of 50°F hotter than the lowest cold leg temperature. 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 (LCO 3.4.6, Note 2.a and LCO 3.4.7, Note 3.a). For this analysis, RCS/SG temperatures of [

]a,c °F,

[

]a,c °F, [

]a,c °F were investigated with PORV setpoints ranging from [

]a,c psig. The results of all HI parametric analyses in Reference 9 are summarized in Table A-2.

It is important to note that LCO 3.4.6, Note 2.b and LCO 3.4.7, Note 3.b, allow an RCP to be started if there is a steam or gas bubble in the pressurizer regardless of the secondary-to-primary T that may exist. HI transients become more severe at higher RCS temperatures and larger secondary-to-primary Ts. Therefore,

- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 4 of 28 to ensure that the design basis HI transient occurring with a 50°F T while water solid remains bounding, a HI transient was analyzed with a steam bubble and the largest postulated secondary to primary T. The Prairie Island technical specification RCP start notes do not define a maximum pressurizer water level for this condition, so it was assumed that having a bubble is defined as having an actual pressurizer level 90%

consistent with LCO 3.4.9. This assumption was confirmed via Reference 5.

[

]a,c The largest secondary-to-primary T assumed in the analysis was 150°F; see Assumption 7. Therefore, this analysis was performed at RCS/SG temperatures of 200/350 °F with the current OPPS PORV setpoint at 500 psig.

Wide Range Pressure and Temperature Uncertainties In accordance with the Reference 10 methodology, pressure and temperature uncertainties were applied to the development of the LTOPS PORV setpoints. Per Reference 5, the wide range temperature and pressure uncertainties are as follows.

Pressure uncertainty

= [

]a,c psi

Temperature uncertainty

= 18.0 °F Pressure Drop between Reactor Vessel and Pressure Transmitter The following P between the reactor vessel mid-plane and pressure transmitter were calculated in Reference 1 and confirmed to remain valid in Reference 5.

For 2 RCPs running for MI

= [

]a,c psi

For 1 RCP running for HI

= [

]a,c psi Pressurizer PORV Characteristics, Stroke, and Delay Times The following summarizes key PORV characteristics used for the analysis.

a. Valve type

= Copes-Vulcan

b. Full open CV (for sub-cooled water discharge)

= [

]a,c gpm / P

c. Valve Opening Stroke Time(1)

= [

]a,c sec (full close to full open)

d. Valve Closing Stroke Time(1,2)

= [

]a,c sec (full open to full close)

e. Signal Delay Time(3)

= [

]a,c sec.

f.

Valve Open/Close Characteristic

= Table 2.1-1 g.

Maximum Pressure Relief Tank Backpressure

= [

]a,c psig

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 5 of 28 Notes:

1.

2.

3.

Table 2.1-1: Pressurizer PORV CV vs. Lift a,c a,c a,c a,c

- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 6 of 28 Appendix G Limits The updated 10 CFR 50, Appendix G P-T limits were calculated by Westinghouse as defined in References 3 and 15. The updated steady state P-T limits will be used in the LTOPS analysis; these are summarized in Table A-3. Per Reference 10, steady-state P-T limits are used for LTOPS setpoint analysis. Note that the limits shown in Table A-3 do not include instrumentation uncertainties; however, these uncertainties are included in the setpoint development as shown in Table A-4.

ASME Code Case N-514 will no longer be applied as it had in References 2 and 8. This code case requires the LTOPS to protect 110% of the KIA based Appendix G limits below the LTOPS enable temperature, which effectively permits a 10% relaxation of the calculated P-T limits. However, since the updated P-T limits were calculated based on KIC reference fracture toughness curve, this relaxation is no longer permitted.

Pressurizer PORV Piping limit In addition to the Appendix G limits, an 800 psig pressure limit is incorporated to address pressurizer PORV piping loading considerations for water relief. This represents the pressure limit at the inlet to the PORV for cold water relief to prevent excessive water-hammer loads in the discharge piping (Reference 10). This limit was applied, but did not govern the determination of the LTOPS PORV setpoint.

RCP No. 1 Seal Differential Pressure (P) Limits

[

]a,c However, per Reference 5, the RCP seal P concern does not apply to the Prairie Island Flowserve RCP seals, which have a seal inlet pressure limit of [

]a,c. Therefore, as is standard for plants with a single LTOPS PORV setpoint, this analysis does not evaluate margin to the RCP seal P limits.

RCS Volumes and Steam Generator Tube Plugging The analyses were performed with RCS volumes obtained from Reference 16. The steam generator model was based on the Framatome Model 56/19 Replacement Steam Generators (RSGs). RCS volumes are input to the model assuming 10% Steam Generator Tube Plugging (SGTP) since smaller RCS volumes are conservative for the analysis. For the HI analysis, the SG heat transfer related inputs (e.g., heat transfer area) are based on 0% SGTP, but the RCS volumes conservatively remain at the values corresponding to 10%

SGTP. Therefore, the analyses were performed to bound any SGTP between 0 and 10%.

2.2 Key Assumptions The following key assumptions were used in the development of the LTOPS PORV setpoints:

1.

It is assumed that the RCS is enclosed by a non-yielding, inelastic boundary. Except for one special case of the HI transient, the pressurizer is assumed to be in a water solid condition with the water at

- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 7 of 28 the same subcooled temperature as the remainder of the RCS. [

]a,c The design basis MI and HI transients are defined in Section 2.1.

2.

Only one PORV was credited to mitigate the low temperature overpressure event to meet the single failure criteria.

3.

[

a,c Therefore, the MI transient is analyzed at the minimum RCS temperature corresponding to the bolt up temperature of 60 °F (Reference 5). The MI case with three charging pumps and a Safety Injection (SI) pump is performed at the minimum RCS temperature of 200 °F for operation with a SI pump enabled (Reference 5).

4.

A single-phase, sub-cooled water discharge through the PORV was assumed.

5.

Letdown flow is conservatively assumed to be isolated during the MI and HI transients.

6.

For the HI transient, both SG secondary sides are conservatively assumed to be at the elevated temperature associated with the defined secondary-to-primary T.

]a,c 7.

The largest secondary-to-primary T assumed in the analysis of the HI event with a steam bubble (LCO 3.4.6, Note 2.b and LCO 3.4.7, Note 3.b) was 150 °F. [

a,c 8.

The HI analysis with a pressurizer steam bubble was performed for the maximum pressurizer water level allowed by technical specification LCO 3.4.9 of 90%. This was confirmed to be acceptable in Reference 5. This was assumed to be the actual pressurizer water level without uncertainties. The appropriate uncertainties should be applied within plant operating procedures to ensure that the actual level does not exceed this analyzed technical specification value.

3.0 Description of Analyses and Evaluations The LTOPS setpoints are determined using the NRC approved methodology in Reference 10. Parametric analyses of the design basis MI and HI transients are performed using the LOFTRAN code. The purpose of these parametric analyses is to generate the transient pressure response data consisting of the PORV setpoint overshoot and undershoots. The LTOPS PORV setpoints are calculated based upon the PORV setpoints overshoot and undershoot data and the LTOPS setpoint acceptance criteria described in Section 4.0.

]

[

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 8 of 28 4.0 Acceptance Criteria The following acceptance criteria from Reference 10 are used to determine the LTOPS 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 acceptance criterion related to the RCP seal P limit is an operational criterion to prevent damage to the RCP seal for Westinghouse analyses as documented in Reference 10. As discussed in Section 2.1 and Reference 5, the RCP seal criterion does not apply to the Prairie Island Flowserve RCP seals. Therefore, consistent with the previous analyses in References 2 and 8, margin to the RCP seal P limits was not evaluated.

5.0 Results and Conclusions 5.1 Transient Analyses 5.1.1 PORV Setpoint Overshoots and Undershoots The pressure overshoot and undershoot are defined as the peak pressure minus the assumed PORV setpoint and the assumed PORV setpoint minus the minimum pressure during the transient, respectively.

MI transients were analyzed at two different flow rates:

[

]a,c gpm, corresponding to three charging pumps injecting at the maximum Hi-Hi flow rate

[

]a,c gpm, corresponding to three charging pumps injecting at the maximum Hi-Hi flow rate and one SI pump at the runout flow Table A-1 shows the summary of RCS pressure data overshoots and undershoots as a function of MI flow rate and PORV setpoint. The values in Table A-1 are calculated at the minimum LTOPS temperature (60 °F) for the lower MI flow rate and 200 °F for the higher MI flow rate.

The overshoot and undershoot pressures resulting from the HI events are shown in Table A-2 as a function of PORV setpoint and RCS/SG temperature with 50°F temperature differential between the SG and RCS.

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 9 of 28 5.1.2 HI Transient with Pressurizer Bubble LCO 3.4.6, Note 2.b and LCO 3.4.7, Note 3.b, allow an RCP to be started if there is a steam or gas bubble in the pressurizer regardless of the secondary-to-primary T that may exist. HI transients become more severe at higher RCS temperatures and larger secondary-to-primary Ts. Therefore, this section documents a special HI case to demonstrate that the design basis HI transient occurring with a 50°F T while water solid (Section 5.1.1 and Table A-2) remains bounding of RCP starts with a pressurizer bubble but larger T.

The analysis was performed with a minimum credible pressurizer bubble corresponding to an actual pressurizer level 90% consistent with LCO 3.4.9 and a secondary to primary T of 150°F; see Assumption 7. Therefore, this analysis was performed at an RCS/SG temperature of [

]a,c °F with the current OPPS PORV setpoint at 500 psig.

This case resulted in a peak pressurizer pressure of [

]a,c psig, which corresponds to an overshoot of

[

]a,c psi above the OPPS PORV setpoint. This case was initiated from an actual RCS temperature of

[

]a,c °F where the adjusted steady state Appendix G P-T limit is [

]a,c psig. It is noted that this peak RCS pressure even remained below the Appendix G limit of [

]a,c psig at the 60 °F boltup temperature.

Therefore, the OPPS PORV setpoint of 500 psig provides adequate protection for the TS LCO 3.4.6, Note 2.b and LCO 3.4.7, Note 3.b RCP start conditions.

Finally, the water solid case at an RCS/SG temperature of [

]a,c °F/[

]a,c °F resulted in a peak pressure of [

]a,c psig, or a [

]a,c psi overshoot. This demonstrates that the pressurizer steam bubble offsets the impact of a larger T and the water solid design cases in Section 5.1.1 remain bounding. Therefore, the water solid design cases remain bounding and OPPS PORV settings calculated based on the water solid design cases (i.e., if different from the current 500 psig setting) will continue to protect the TS LCO 3.4.6, Note 2.b and LCO 3.4.7, Note 3.b RCP start conditions.

Key parameter responses for this transient are provided in Figures B-1 through B-5.

5.2 PORV Setpoints Determination Using the results of the LTOPS design basis MI and HI parametric transient analyses from Tables A-1 and A-2, the LTOPS maximum allowable PORV setpoints valid up to 54 EFPY are determined. The PORV setpoint of 500 psig from Reference 2 was then investigated to ensure that it remained bounding. The transients investigated remained the same from the analysis of record, Reference 2. However, initial analysis with the prior transient definitions showed that the OPPS PORV setpoint would need to be reduced to protect the revised Appendix G limits. Therefore, the following changes from the prior analysis have been made to recover margin and maintain the current OPPS setting. Xcel concurred with these changes via Reference 5:

[

] a,c

]a,c

[

]

a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 10 of 28 A summary of the maximum allowable LTOPS PORV setpoints calculations and associated limits for the MI transient is shown in Table A-5 and for the HI transient in Table A-6. The maximum allowable PORV setpoints for the MI and HI transients are plotted as a function of indicated RCS temperature in Figure A-1.

The final maximum allowable PORV setpoints are determined such that it bounds both the MI and HI transient maximum allowable PORV setpoints. The final maximum allowable LTOPS PORV setpoints are shown in Table A-7 and Figure A-1.

The current single setpoint of 500 psig is shown via the black line on Figure A-1 marked as Current Setpoint. As can be seen from the gray line on Figure A-1 marked as Max Allowable, the current OPPS setpoint of 500 psig remains bounding and can be maintained through 54 EFPY at Units 1 and 2. If Xcel wanted to pursue implementation of a variable OPPS PORV setting, it could be based on the values in Table A-7 and the margin between these settings and the current setting of 500 psig would be recovered as operating margin.

The evaluation showed that the current PTLR OPPS setting of 500 psig maintains at least [

]a,c psi of margin to the maximum allowable settings calculated throughout the range of LTOP applicability. Therefore, the current LTOPS settings shown in Figure A-1 are bounding and can be maintained to 54 EFPY for Prairie Island Units 1 and 2.

5.3 OPPS PORV Cycling Pressurizer PORV cycling analyses have been performed in References 2 and 11 to support backup air supply sizing evaluations. Since the HI transient is self-terminating when thermal equilibrium is reached, the MI transient is limiting for PORV cycling. As mentioned in Reference 2, the PORV cycling frequency depends on the following key parameters:

[

] a,c

[

] a,c

[

] a,c

[

] a,c

[

] a,c Therefore, the previous OPPS PORV cycling analyses remain valid.

5.4 LTOPS Arming / Enable Temperature The LTOPS Enable temperature was calculated in Reference 3 using the methods of ASME Code Case N-641 to be 206 °F (without uncertainty). With the temperature uncertainty of 18 °F applied, the minimum arming/enable temperature is 224 °F. Per Reference 7, the current manual arming temperature at Prairie Island Units 1 and 2 is 310 °F and it was requested in Reference 5 to investigate reducing this enable temperature to 290 °F. The reduced enable temperature of 290 °F remains conservatively higher than the minimum required enable temperature of 224 °F and is therefore acceptable for reactor vessel integrity.

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 11 of 28

[

a,c

[

a,c

[

] a,c Based on the above evaluation, an OPPS enable temperature of 290 °F is acceptable for up to 54 EFPY and this meets the requirements of the ASME code, NUREG-0800 BTP 5-2, and 10 CFR 50 Appendix G.

[

a,c Therefore, it is recommended that Prairie Island consider whether changes to the plant operating procedures that govern water solid operations are necessary to accommodate the reduced OPPS enable temperature.

]

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 12 of 28 6.0 References

1.

2.

3.

4.

5.

6.

Technical Specifications LCO 3.14.12 and 3.4.13, Northern States Power Company, Docket No. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Unit 1 Renewed Facility Operating License and Prairie Island Nuclear Generating Plant, Unit 2 Renewed Facility Operating License, Amendments 237 and 225.

7.

Prairie Island Nuclear Generating Plant Units One and Two Pressure and Temperature Limits Report, Revision 6, August 2, 2019.

8.

9.

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

11.

12.

13.

14.

15.

a,c a,c a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 13 of 28

16.

a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 14 of 28 Appendix A Table A-1: Prairie Island Units 1 and 2 Mass Injection Pressure Overshoots/Undershoots Summary a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 15 of 28 Table A-2: Prairie Island Units 1 and 2 Heat Injection Pressure Overshoots/Undershoots Summary a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 16 of 28 a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 17 of 28 a,c

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- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 19 of 28 Table A-4: Adjusted Appendix G Limits for Prairie Island Units 1 and 2 at up to 54 EFPY a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 20 of 28 a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 21 of 28 Table A-5: Prairie Island Units 1 and 2 Maximum Allowable Setpoint Determination for the MI Transient a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 22 of 28 a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 23 of 28 Table A-6: Prairie Island Units 1 and 2 Maximum Allowable Setpoint Determination for the HI Transient a,c

- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 25 of 28 Figure A-1: Prairie Island Units 1 and 2 Combined Maximum Allowable PORV Setpoint (includes Pressure and Temperature uncertainties) a,c

- - This record was final approved on 8/3/2022, 12:00:05 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 26 of 28 Appendix B Figure B-1: RCS Pressure Figure B-2: RCS Temperature a,c a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 27 of 28 Figure B-3: Pressurizer Water Volume Figure B-4: RCS Loop Flow a,c a,c

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Westinghouse Non-Proprietary Class 3 NSPM-LTP-TR-AA-000001-NP, Rev. 1 Page 28 of 28 Figure B-5: Liquid/Steam Relief Flow a,c

ENCLOSURE 4 Westinghouse Affidavit Application for Withholding Proprietary Information from Public Disclosure 3 pages follow

- - This record was final approved on 8/2/2022, 6:25:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-034 Page 1 of 3 Commonwealth of Pennsylvania:

County of Butler:

(1)

I, Camille Zozula, Manager, Regulatory Compliance and Corporate Licensing, 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 NSPM-LTP-TR-AA-000001-P be withheld from public disclosure under 10 CFR 2.390.

(3)

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

(4)

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

(i)

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

(ii)

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

(iii)

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

- - This record was final approved on 8/2/2022, 6:25:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-034 Page 2 of 3 (5)

Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

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

(b)

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

(c)

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

(d)

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

(e)

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

(f)

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

(6)

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

- - This record was final approved on 8/2/2022, 6:25:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-034 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 8/2/2022 Signed electronically by Camille Zozula Executed on: 8/2/2022 Signed electronically by

v t e Letter Sequence Request
EPID:L-2022-LLA-0184, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Supplement Administration Withholding Request Acceptance Questions Answers 20 December 2023 Results Approval Other: ML23348A347, ML24271A292
MONTHYEAR2023-01-26 [Table View]

L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

Text

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

Response

Response

Response

Response

Response

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

2.1 INTRODUCTION

SUMMARY

2.1 INTRODUCTION

Response

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