L-24-189, License Renewal Application for Revision O - Supplement 1

License Renewal Application for Revision O - Supplement 1

ML24220A270
Person / Time
Site:	Perry
Issue date:	08/07/2024
From:	Penfield R Vistra Operations Company
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-24-189
Download: ML24220A270 (1)

Text

Perry Nuclear Power Plant Rod L. Penfield Site Vice President 10 Center Road Perry, Ohio 44081

L-24-189 August?,2024 10 CFR 54

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant, Unit No. 1 Docket No. 50-440, License No. NPF-58 License Renewal Application for the Perry Nuclear Power Plant Revision O - Supplement 1 (Non Proprietary)

REFERENCES:

1. Letter L-23-146, from Rod L. Penfield to the Nuclear Regulatory Commission, dated July 3, 2023, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, (ADAMS Accession No. ML23184A081)

2. Nuclear Regulatory Commission Letter from Lauren K. Gibson to Rod L. Penfield, Perry Nuclear Power Plant, Unit No. 1 dated September 25, 2023 - Aging Management Audit Plan Regarding the License Renewal Application Review, (ADAMS Accession No. ML23261B019)

3. Letter L-24-109, from Rod L. Penfield to the Nuclear Regulatory Commission, dated May 30, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 1 (ADAMS Accession No. ML24151A637).

4. Letter L-24-110, from Rod L. Penfield to the Nuclear Regulatory Commission, dated July 3, 2024, submitting 10 CFR 54.21 (b) Annual Amendment to the Perry Nuclear Power Plant License Renewal Application (ADAMS Accession No. ML24185A092)

6555 SIERRA DRIVE IRVING. TEXAS 75039 o 214-812-4600 VISTRACORP.COM Perry Nuclear Power Plant L-24-189 Page 2 of 2

In Reference 1, Energy Harbor Nuclear Corp, submitted a license renewal application (LRA) for the Facility Operating License for Perry Nuclear Power Plant, Unit No. 1 (PNPP). Since submittal of the*

LRA, the PNPP Facility Operating License has been transferred to Vistra Operations Company LLC (VistraOps) per conforming license Amendment 203 and the license transfer transaction was closed on March 1, 2024 (EPID L-2024-LLM-0000). The license transfer changes impacting the PNPP LRA are documented in Reference 4, which is the annual amendment required by 10 CFR 54.21(b).

During the Nuclear Regulatory Commission (NRC) staff's aging management audit of the PNPP LRA (Reference 2), PNPP Staff agreed to supplement the LRA with clarifying information which has led to several LRA supplements. In Reference 3, VistraOps submitted the first supplement (Supplement 1) to the PNPP LRA that provided clarifying information for the electrical and Time Limited Aging Analysis (TLAA) portion of the LRA addressed via the audit and breakout discussions with the NRC auditors.

The attachment to this letter provides submittal of Supplement 1 with Attachment 9 updated to address proprietary information.

For ease of reference, an index listing the supplemented LRA sections and tables and the associated attachment is provided. In each attachment the changes are described along with the affected section(s) and page number(s) of the LRA (Reference 1 ). For clarity, supplemented information is indicated by, red, bolded and underlined text and text to be deleted is indicated by strikethrough. Revisions to LRA tables may be shown by providing excerpts from each affected table.

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Mark Bensi, PNPP License Renewal Project Manager at (440) 280-6179 or via email at Mark.Bensi@vistracorp.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 7, 2024.

Sincerely, t(}d,;;2;*.

Rod L. Penfield

Attachments:

PNPP LRA Supplement 1 (Non-Proprietary) - Attachments for Electrical and TLAA Sections and Tables

cc: NRC Region Ill Administrator NRC Resident Inspector NRR Project Manager Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary)

L-24-189 Attachments Index Page 1 of 2

PNPP LRA Supplement 1 (Non-Proprietary)

Attachments Index for Electrical and TLAA Sections and Tables

Attachment LRA Section, Table No. or Appendix Subject Inquiry No.

Supplemented Electrical Issues 1 Section 2.1.1.3.5 Station Blackout S&S 2.1.1.3.5-04 S&S 2.1.1.3.5-05 2 Section 2.1.2.3 Scope of Electrical S&S 2.1.2.3-01 Components 3 Table 3.6.2 Electrical Commodities & TRP-057-01 EQ Program 4 Appendix B Fuse Holders Program S&S B.2.24-05 Section B.2.24 Time Limited Aging Analysis (TLAA) Issues 5 Section 4.1-3 TLAA Analysis Results TRP-063-03 Table 4.1-2 Changes TRP-116.11-01 6 Section 4.2.1 BWRVIP benchmarking TRP-059-1-01 guidance 7 Section 4.2.2 Table revision to reflect TRP-059.3-02 Table 4.2-2 rounding for Heat No.

8 Section 4.2.3 Table revision to reflect TRP-059.3-02 Table 4.2-3 rounding for Heat No.

9 Section 4.2.5 Table clarification margin TRP-059.5-01 Table 4.2-5 TRP-059.5-02 10 Section 4.2.6 Main Steam Line Break TRP-059.6-01 transient clarification 11 Section 4.3.2 Provide additional detail for TRP-060.2-03-01 Table 4.3-3A 60 cycle projections for non-Class 1 systems 12 Section 4.3.3 Clarification for TRP-060.3-02 NUREG/CR-6260 guidance TRP-060.3-03 13 Section 4.3.3 Clarification for Table 4.3-5 NUREG/CR-6260 guidance TRP-060.3-02 14 Section 4.3.5 Clarification for HELB TRP-060.5-01 Section 4.3.6 Location Determination TLAA 15 Section 4.5.3 Clarification for TRP-063-03 Containment Piping Penetration Bellows 16 Section 4.6.8 Clarification for Silicone Applicant Initiated Sealant Testing Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary)

L-24-189 Attachments Index Page 2 of 2

Attachment LRA Section, Table No. or Appendix Subject Inquiry No.

Supplemented 17 Section 4.6.11 Deletion of Allowable Stress TRP-116.11-01 Allowable Stress Analysis of BOP ASME CODE Class 1, 2 and 3 Components due to redundancy 18 Section 4.7 Added reference TRP-059-1-01

19 Appendix A Correction of typographical Applicant Initiated Section A.2.2.4 error 20 Appendix B Correction of typographical Applicant Initiated Section B.2.14 error 21 Appendix C Clarification of how TRP-009-01 BWRVIP documents are referenced

Key for Attachments:

Frequently Used Acronyms:

LRA = License Renewal Application

TRP = Technical Review Package from NRC aging management audit inquiry

S&S = NRC Scoping and Screening audit inquiry

For clarity, supplemented information is indicated by, red, bolded and underlined text and text to be deleted is indicated by strikethrough. Revisions to LRA tables may be shown by providing excerpts from each affected table.

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LRA Section: Section 2.1.1.3.5

LRA Page Number(s): Page 2.1-12

References:

S&S 2.1.1.3.5-04 and S&S 2.1.1.3.5-05

Description of Change: LRA section 2.1.1.3.5 identified that the plant boundary for SBO recovery is extended to the first interconnective devices [switchyard breakers] and the step-up transformer. The NRC staff noted there is no step-up transformer identified on the provided boundary drawing nor are the circuit breakers highlighted.

LRA Section 2.1.1.3.5 is revised to clarify the step-up station transformers are also called the startup transformers. LRA Figure 2.1-1 Electrical One Line Diagram 13.8 kV and 4.16 kV is added to depict the station blackout recovery boundary, among other details.

The LRA credited manual circuit breaker operation for station blackout recovery and did not include the 125 VDC control cables for the breakers. LRA Section 2.1.1.3.5 is revised to include the control cable for switchyard breakers and their support structures.

PNPP LRA Section 2.1.1.3.5, Page 2.1-12 is revised as follows:

2.1.1.3.5 Station Blackout (10 CFR 50.63)

10 CFR 50.63 [Reference 1.3-7], requires that each light-water-cooled nuclear power plant be able to withstand, for a specified duration, and recover from a station blackout (SBO).

An SBO is the loss of offsite and onsite AC electric power to the essential and nonessential switchgear buses in a nuclear power plant. It does not include the loss of AC power fed from inverters powered by station batteries or by alternate AC sources. The objective of this requirement is to assure that nuclear power plants can withstand an SBO while maintaining adequate reactor core cooling and containment integrity for the specified duration.

Appendix 15H of the UFSAR describes the licensing bases for SBO at the Perry Nuclear Power Plant (PNPP). PNPP has developed a four-hour coping analysis to address the requirements of 10 CFR 50.63.

Based on the PNPP current licensing bases for SBO, structure and system intended functions performed in support of 10 CFR 50.63 requirements were determined. The results of this determination are provided for mechanical systems in Section 2.3, for structures in Section 2.4, and electrical commodities in Section 2.5.

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The For Station Blackout Recovery, the license renewal scoping boundary for SBO is the step-up station transformer for Unit 1 and Unit 2. For Station Blackout Recovery, the boundary is extended to the first interconnection device that would restore offsite power to the main switchyard busses and then to the step-up station transformer (startup transformers). At PNPP, the boundary with the offsite transmission system has been defined at the 345kV switchyard circuit breakers: breakers S-612, S-620, S-621, S-650, S-652, S-660 and S-661. Refer to sScoping boundary drawing 206-0010 LR(Figure 2.1-1) shows the Station Blackout Recovery boundary in red. This boundary definition is consistent with NUREG-1800 [Reference 1.3-9], Section 2.1.3.1.

The Transmission Yard Control House (and associated EIC components) is not credited for offsite power supply reliability or station blackout recovery and is not in the scope of license renewal. The 125 VDC control circuits for the switchyard boundary breakers, and their protective structures are included in scope and subject to aging management review.

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LRA Section: Section 2.1.2.3

LRA Page Number(s): Page 2.1-18

Reference:

S&S 2.1.2.3-01

Description of Change: LRA Section 2.1.2.3 addresses screening of electrical systems, components and commodities. It was not clear this included instrument and control electrical systems, components and commodities. Therefore, LRA Section 2.1.2.3 is revised to clarify that instrument and control (I&C) systems, components and commodities are included and that all commodity groups at the Perry Nuclear Power Plant (PNPP) are grouped consistent with NEI 95-10 and NUREG-1800.

PNPP LRA Section 2.1.2.3, Page 2.1-18 is revised as follows:

2.1.2.3 Screening of Electrical and I&C Systems

Electrical All electrical and I&C systems, and electrical and I&C components within mechanical systems, did not require further system evaluations to determine which components were required to perform or support the identified intended functions. A bounding scoping approach is used for electrical and I&C equipment. All electrical and I&C components within in-scope systems were included within the scope of license renewal. In-scope electrical components were placed into commodity groups and were evaluated as commodities during the screening process identified from a review of electrical systems within the scope of 10 CFR 54, controlled electrical drawings, the SAP functional location database, and interface with the mechanical and structural screening process. This commodity-based approach, whereby component types with similar design and/or functional characteristics are grouped together, is consistent with the guidelines from NEI 95-10 and Table 2.1-5 of NUREG-1800.

The screening phase for electrical and I&C components starts by comparing in-scope commodity types to the commodity types listed in Appendix B of NEI 95-10 [Reference 1.3-3]. NEI 95-10 provides guidance for determining whether the commodities are active or passive. Active commodities are screened out.

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Short-lived components or commodities do not require aging management review, so the short-lived components are screened out. 10 CFR 54.21 (a)(1)(ii) [Reference 1.3-1]

allows the exclusion of those commodities that are subject to replacement based on a qualified life or specified time period. Electrical and I&C commodities included in the Environmental Qualification (EQ) Program meet that exclusion criterion. This is because those components and commodities have defined qualified lives and are replaced prior to the expiration of their qualified lives.

Section 2.5 presents the results of the screening process for electrical and I&C systems.

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LRA Section: Table 3.6.2

LRA Page Number(s): Page 3.6-25 (Add Page 3.6-25A)

Reference:

TRP-057-01

Description or Changes: It was noted that, while Environmental Qualification (EQ) is a Time Limited Aging Analysis (TLAA), an aging management program is described in LRA Appendix A and B Sections A.1.17, A.2.4 and B.2.17. The EQ components are not included in Table 3.6.2, Electrical Commodities Summary of Aging Management Evaluation. Two lines for electrical equipment subject to 10 CFR 50.49 have been added to Table 3.6.2.

Identified change:

PNPP LRA Section Table 3.6.2, Page 3.6-25 is revised as follows: (see next page)

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LRA Page Number(s): Page B-79

Reference:

S&S B.2.24-05

Description of Change: It was noted LRA Appendix B, Section B.2.24 was not consistent with NUREG 1801 AMP XI.E5 in that contact resistance testing was not described in the AMP and that the industry operating experience had not been cited.

LRA Appendix B, Section B.2.24 is revised to credit thermography, contact resistance testing, or other appropriate test methods. In addition, the operating experience referenced in NUREG-1801 is added and a typographical error on the operating experience review timeframe is corrected.

PNPP LRA Section B.2.24, Page B-79 is revised as follows:

B.2.24 FUSE HOLDERS PROGRAM

Program Description

The Fuse Holders Program is a new Condition Monitoring Program. The program provides reasonable assurance that the intended functions of the metallic clamps of fuse holders located outside of active devices are maintained consistent with the current licensing basis through the period of extended operation. Fuse holders located inside an active device are not within the scope of this program. Fuse holders subject to increased resistance of connection due to chemical contamination, corrosion, and oxidation or fatigue caused by ohmic heating, thermal cycling or electrical transients will be tested, by a proven test methodologythermography, contact resistance testing, or other appropriate test methods, at least once every 10 years to provide an indication of the condition of the metallic clamps of the fuse holders.

The first tests for license renewal will be completed no later than six months prior to the period of extended operation.

NUREG-1801 Consistency

The Fuse Holders Program is a new PNPP program that is consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.E5, Fuse Holders.

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Exceptions to NUREG-1801

None

Enhancements

None

Operating Experience

Industry operating experience and guidance as documented in NUREG-1760, IEEE Std. 1205-2000, and NRC Information Notices 86-87, 87-42 and 91-78 has shown that loosening of fuse holders and corrosion of fuse clips are aging mechanisms that, if left unmanaged, could lead to a loss of electrical continuity if left unmanaged. A review of plant-specific operating experience has found no evidence of thermal anomalies in annual tests of fuse holders outside active devices from 2003 to 2021 2013 to 2023.

Conclusion

The implementation of the Fuse Holders Program with enhancement, provides reasonable assurance that the aging effects will be managed such that components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

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Attachment No. 5

LRA Section: Table 4.1-2

LRA Page Number(s): Page 4.1-7 and Page 4.1-8

References:

TRP-063-03 and TRP-116.11-01

Description of Change: Changes to LRA Table 4.1-2 are made to reflect the results of TLAA analyses that have been updated from NRC breakout inspection questions and responses:

In Attachment 15 to this supplement, LRA section 4.5.3 is being updated to reflect that the Containment Piping Penetration Bellows TLAA will be managed per 10 CFR 54.21(c)(1)(iii).

In Attachment 17 to this supplement, the identified TLAA in LRA Section 4.6.11, Allowable Stress Analysis of BOP ASME CODE Class 1, 2 and 3 Components, was found to be redundant to existing sections of the LRA addressing the same TLAAs:

o Class 1 piping and components are addressed in LRA Section 4.3.1.

o Class 2 and 3 piping and components are addressed in LRA Section 4.3.2.

In order to eliminate potential confusion, LRA Section 4.6.11 of the LRA will be deleted.

PNPP LRA Table 4.1.2, Pages 4.1-7 and 4.1-8 are revised as follows: (see following pages)

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Table 4.1 TLAA Results Summary Subject TLAA SectioDisposition n

Neutron Fluence 4.2.1 TLAA projected to the end of the PEO per 10 CFR 54.21(c)(1)(ii).

Upper Shelf Energy 4.2.2 TLAA projected to the end of the PEO per 10 CFR 54.21(c)(1)(ii).

Reactor Adjusted Reference 4.2.3 TLAA projected to the end Pressure Temperature Analyses of the PEO per Vessel 10 CFR 54.21(c)(1)(ii).

Neutron Pressure-Temperature (P-T) 4.2.4 TLAA will be managed Embrittlement Limits per 10 CFR 54.21(c)(1)(iii).

RPV Shell Weld Failure 4.2.5 TLAA projected to the end Probability Assessment of the PEO per Analyses 10 CFR 54.21(c)(1)(ii).

RPV Reflood Thermal Shock 4.2.6 TLAA projected to the end of the PEO per 10 CFR 54.21(c)(1)(ii).

Reactor Pressure Vessel 4.3.1.1 TLAA will be managed per 10 CFR 54.21(c)(1)(iii).

Class 1 Piping 4.3.1.2 TLAA will be managed per 10 CFR 54.21(c)(1)(iii).

Non-Class 1 Fatigue 4.3.2 TLAA will remain valid for Metal Fatigue the PEO per 10 CFR 54.21(c)(1)(i).

Environmental Fatigue 4.3.3 TLAA will be managed per 10 CFR 54.21(c)(1)(iii).

Reactor Vessel Internals 4.3.4 TLAA will be managed Fatigue per 10 CFR 54.21(c)(1)(iii).

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Table 4.1 TLAA Results Summary Subject TLAA SectioDisposition n

Intermediate HELB location 4.3.5 TLAA will be managed Determination per 10 CFR 54.21(c)(1)(iii).

Environmental EQ Evaluations 4.4 TLAA will be managed Qualification per for Electrical 10 CFR 54.21(c)(1)(iii).

Equipment Containment Vessel 4.5.1 TLAA will be managed per 10 CFR 54.21(c)(1)(iii).

Containment Piping 4.5.2 TLAA will be managed Penetrations per 10 CFR 54.21(c)(1)(iii).

Containment Piping 4.5.3 TLAA will be managed Penetration Bellows per 10 CFR 54.21(c)(1)(iii).

TLAA will remain valid for the PEO per 10 CFR 54.21(c)(1)(i).

Other Crane Load Cycles 4.6.1 TLAA will remain valid for Potential the PEO per Plant-Specific 10 CFR 54.21(c)(1)(i).

TLAAs Main Steam Line Flow 4.6.2 TLAA projected to the end Restrictors Erosion Analysis of the PEO per 10 CFR 54.21(c)(1)(ii).

Reduction of Fracture 4.6.3 TLAA will be managed Toughness for the Reactor per Vessel Internals 10 CFR 54.21(c)(1)(iii).

Fatigue Analysis -- 4.6.4 TLAA will remain valid for Earthquake Cyclic Loading the PEO per 10 CFR 54.21(c)(1)(i).

Fatigue due to Partial 4.6.5 TLAA will be managed Feedwater Heating per 10 CFR 54.21(c)(1)(iii).

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Table 4.1 TLAA Results Summary Subject TLAA SectioDisposition n

Fatigue due to Single 4.6.6 TLAA projected to the end Recirculation Loop Operation of the PEO per 10 CFR 54.21(c)(1)(ii).

Steam Piping Erosion 4.6.7 TLAA will be managed per 10 CFR 54.21(c)(1)(iii).

Silicone Sealant in ESF 4.6.8 TLAA will be managed HVAC per 10 CFR 54.21(c)(1)(iii).

Top Guide Grid Beam 4.6.9 TLAA will be managed Neutron Fluence per 10 CFR 54.21(c)(1)(iii).

Jet Pump Fatigue Analysis 4.6.10 TLAA will be managed per 10 CFR 54.21(c)(1)(iii).

Allowable Stress Analysis of 4.6.11 TLAA will remain valid for BOP ASME Code Class 1, 2 the PEO per 10 CFR and 3 Components 54.21(c)(1)(i).

RV Annealing 4.6.12 TLAA will remain valid for the PEO per 10 CFR 54.21(c)(1)(i).

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LRA Section: Section 4.2.1

LRA Page Number(s): Pages 4.2-2 through Page 4.2-4

Reference:

TRP-059-1-01

Description of Change: The LRA section is to be revised to clarify that PNPP has met benchmarking guidance in the NRC SERs for the applicable BWRVIP documents to demonstrate the applicability of the RAMA fluence methodology to the PNPP reactor vessel and internals.

PNPP LRA Table 4.2.1, Page 4.2-3, is revised as follows:

4.2.1 Neutron Fluence

TLAA

Description:

Neutron fluence is the term used to represent the cumulative number of neutrons per square centimeter that contact the reactor vessel shell and its internal components over a given period of time. The fluence projections that quantify the number of neutrons that contact these surfaces have been used as inputs to the neutron embrittlement analyses that evaluate the loss of fracture toughness aging effect resulting from neutron fluence.

Fluence projections were performed to predict the neutron fluence expected to occur during 32 Effective Full Power Years (EFPY) of plant operation. The 32 EFPY fluence calculations were submitted in a letter dated June 4, 2002 [Reference 4.7-2] and approved by the NRC in a letter dated April 29, 2003 [Reference 4.7-3]. At the time the projections were prepared, 32 EFPY was considered to represent the amount of power to be generated over 40 years of plant operation, assuming a 40-year average capacity factor of 80 percent. These fluence projections have been identified as TLAAs requiring evaluation for the period of extended operation.

USAR Sections 4.3.2.8 and 5.3.1.6.2 provide evaluations of the effect of neutron fluence on reactor vessel materials. These evaluations are considered to be TLAAs.

BWRVIP-74-A, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal, [Reference 4.7-26] identified the evaluation of the loss of reactor vessel fracture toughness due to neutron fluence as a TLAA.

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TLAA Evaluation:

EFPY projections:

End-of-life fluence is based on a projected value of EFPY over the licensed life of the plant. The full power operating license (FPOL) for PNPP Unit 1 was issued in November 1986. PNPP Unit 1 was originally licensed for a maximum thermal power of 3579 MWt.

License Amendment No. 112, issued June 1, 2000, increased the maximum thermal power to 3758 MWt through a 5% thermal power uprate. License Amendment No. 191, issued on October 8, 2020, revised the expiration date of PNPP's FPOL such that it would expire 40 years from the date of issuance of the FPOL, as opposed to 40 years from the date of issuance of the fuel loading and low-power testing license. These changes to the Operating License are included in the fluence projections.

Operating cycle 18 was completed in the spring of 2021 with the accrued effective full power years (EFPY) of 27.3, representing approximately 35 years of operation. The projected EFPY though the end of the period of extended operation using a 96 percent average capacity factor (assumes 100 percent capacity factor between refueling outages and 30-day refueling outages every 2 years) is less than 54 EFPY at end of 60 years of operation.

Fluence projections:

The fluence values provided in this section were calculated using the Radiation Analysis Modeling Application (RAMA) Fluence Methodology. RAMA was developed for the Electric Power Research Institute and the Boiling Water Reactor Vessel and Internals Project. The NRC has reviewed and approved RAMA for BWR reactor pressure vessel (RPV) fluence predictions by letter dated February 7, 2008 [Reference 4.7-4]. Use of this methodology for evaluations of fluence for PNPP was performed in accordance with guidelines presented in NRC Regulatory Guide 1.190, [Reference 4.7-5]. In compliance with these guidelines, comparisons to surveillance capsule flux wire and dosimetry measurements were performed to determine the accuracy of the RPV fluence model.

An uncertainty analysis was also performed to determine if a statistical bias exists in the model. It was determined that the PNPP model does not have a statistical bias and that the best-estimate fluence is suitable for use in evaluating the effects of embrittlement on RPV material as specified in 10 CFR 50 Appendix G [Reference 4.7-6] and NRC Regulatory Guide 1.99, Revision 2 [Reference 4.7-7]. Therefore, PNPP has satisfied the benchmarking guidance in the NRC SERs for the BWRVIP-114, -115, -117, and Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary)

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-121. [Reference 4.7-13] and for BWRVIP-145 to demonstrate the applicability of the RAMA fluence methodology to the PNPP reactor vessel and internals.

Fast neutron fluence evaluation was performed for the RPV based on operating data through cycle 14. Fluence was calculated at EOC 14 (20.0 EFPY) and projected to 54 EFPY. In LRA Table 4.2-1, PNPP RPV Beltline Fluence Data for 54 EFPY, fast neutron fluence for energy >1.0 MeV is reported for the RPV plates, welds and nozzles throughout the RPV beltline region at the interface of the base metal and cladding, hereafter denoted as the 0t location of the RPV wall. Fluence attenuations are performed through the RPV wall to the 1/4t locations using the displacement per atom (DPA) attenuation method prescribed in NRC Regulatory Guide 1.99, Revision 2.

Fluence values that exceed the threshold value of 1.0E+17 n/cm2 for 54 EFPY define the RPV beltline for the period of extended operation. The maximum fluence value is 4.79E+18 n/cm2 for the lower-intermediate shell plate at the 0t location.

Neutron fluence analysis valid for 54 EFPY have been prepared for the reactor vessel beltline materials and bound the projected EFPY value for 60-years of operation.

Therefore, the neutron fluence analysis has been projected to the end of the period of extended operation.

Disposition: 10 CFR 54.21(c)(1)(ii) The neutron fluence analysis has been projected to the end of the period of extended operation.

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LRA Section: Section 4.2.2, Table 4.2-2

LRA Page Number(s): Pages 4.2-7 and Page 4.2-8

Reference:

TRP-059.3-02

Description of Change: The LRA table is to be revised to make the rounding of the Cu (wt %) for Heat No. 5P6214B to two decimal places consistent throughout the table.

PNPP LRA Table 4.2-2, Pages 4.2-7 and 4.2-8, is revised as follows: (see following pages)

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LRA Section: Section 4.2.3, Table 4.2-3

LRA Page Number(s): Pages 4.2-11 through 4.2-13

Reference:

TRP-059.3-02

Description of Change: Table 4.2-3 is revised to make the rounding of the Cu (wt %)

for Heat No. 5P6214B to two decimal places consistent throughout the table.

PNPP LRA Table 4.2-3, Pages 4.2-12 is revised as follows: (see following pages)

a

a a

a a

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LRA Section: Section 4.2.5, Table 4.2-5

LRA Page Number(s): Page 4.2-16

Reference:

TRP-059.5-01 and TRP-059.5-02

Description of Change: Table 4.2-5 is revised to clarify that once the RPV surveillance data is considered, the Lower-Intermediate Shell Plate 22-1-1, Heat No.

C2557-1 becomes the limiting plate, to clarify that End of Interval (EOI) RTMAX (°F) does not include margin, and to remove the margin inappropriately included in the EOI RTMAX

(°F) values. Proprietary information of Section 4 that have been removed are indicated by a set of open and closed double bold brackets as shown here ((.

PNPP LRA Table 4.2-5, Pages 4.2-16 is revised as follows: (see next page)

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Table 4.2-5 PNPP RPV Material Adjusted Reference Temperature for 54-EFPY

Parameter Limiting Limiting Limiting Plate Circumferential Weld Axial Weld Component No. Lower-Intermediate Lower Int/Lower Lower Intermediate Shell Plate 22-1-1 Shell Girth Weld Shell Axial Weld Lower-Intermediate Shell Plate AB BD, BF

22-1-3 Heat / Lot Identification C2557-1 A1155-1 4P7216 (single wire) 627260 Number 4P7216 (tandem wire)

Copper Content (wt. %) 0.05 0.06 0.03 0.06

Nickel Content (wt. %) 0.63 0.79 (single wire) 1.08

0.81 (tandem wire)

Chemistry Factor (CF) 37 36 41 82 (°F)

54-EFPY 4.79 E+18 4.85 x E+17 3.31 x E+18 (BD)

EOI Neutron Fluence 3.15 x E+18 (BF) (f) (n/cm2)

RTNDT(U) (°F) 10 -10 -20 -30

EOI RTNDT (°F) 29.0 29.4 11.8 57.1

EOI RTMAX (°F) = 39 48.8 -8.2 3.6 27.1 83.1 RTNDT(U) + RTNDT EOI RTMAX (°F)

(( }} (( }} (( }} (( }}

Limiting RTMAX (°F)

EOI RTMAX < Limiting Yes Yes Yes RTMAX?

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 10 Page 1 of 4

0

LRA Section: 4.2.6

LRA Page Number(s): Page 4.2-17 through Page 4.2-19

Reference:

TRP-059.6-01

Description of Change: LRA Section 4.2.6 is revised to clarify that the steam line break is the controlling transient to verify the integrity of the reactor vessel during a reflood event; therefore, the use of the more realistic 0.052t flaw in the recirculation line evaluation is acceptable.

PNPP LRA Section 4.2.6, Page 4.2-18, is revised as follows:

4.2.6 RPV Reflood Thermal Shock

TLAA

Description:

10CFR50 Appendix A, General Design Criterion (GDC) 31, requires that the reactor coolant pressure boundary (RCPB) of a light water reactor (LWR) be designed such that it possesses adequate margin against non-ductile failure for all postulated conditions.

For General Electric (GE) designed Boiling Water Reactors (BWRs) this requirement has been demonstrated through development of Pressure-Temperature Limit Curves (P-T curves) and reference to generic analyses [References 4.7-14 and 4.7-15], which address the limiting Loss of Coolant Accident (LOCA) event. The acceptance criterion used in these analyses is that the crack driving force for postulated flaws in the reactor pressure vessel (RPV), KI present during the bounding Emergency or Faulted condition (Service Level C and D), is less than the limiting material resistance to fracture, KIC, applicable during the event. The analysis performed to address service level C/D conditions is often referred to as the RPV reflood thermal shock analysis.

Since the initial RPV reflood thermal shock analyses are based upon fluence values associated with 40 years of operation, they constitute TLAAs requiring evaluation for 54 EFPY through the period of extended operation.

TLAA Evaluation:

The updated PNPP RPV reflood thermal shock analysis considered plant operation through the period of extended operation (PEO) in order to demonstrate that adequate Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 10 Page 2 of 4

margin against non-ductile failure of the PNPP RPV exists as required by 10CFR50, Appendix A, GDC 31.

The generic analyses [References 4.7-14 and 4.7-15] only addressed this requirement for the RPV shell plates and welds since, at the time these analyses were prepared, the nozzles were not expected to receive sufficient irradiation during the life of the plant to require inclusion within the beltline. Due to increased capacity factors and power uprates, the N6 LPCI nozzles and N12 water level instrument nozzles have been added to the beltline and have been included in the updated evaluation to demonstrate adequate margin against non-ductile failure for all beltline materials during the PEO.

Beltline Shell Materials

The updated PNPP RPV reflood thermal shock analysis demonstrated that all beltline materials in the RPV will satisfy the acceptance criteria for postulated flaw sizes less than or equal to the flaws acceptable, without evaluation, in ASME Section XI IWB-3500 and considering operation through the end of the PEO. The beltline shell materials (plates and welds) were evaluated, with the main findings summarized below.

The limiting adjusted reference temperature (ART) at the RPV inside surface for beltline plates and welds at 54 EFPY is 83.1°F, corresponding to weld heat number 627260 (lower intermediate shell axial weld BD, BF). The methodology for development of ART is addressed in LRA Section 4.2.3. The limiting ART of 83.1°F was used to establish fracture toughness allowables for the RPV reflood thermal shock analysis.

To ensure that the RPV can satisfactorily withstand the effects of the reflood thermal shock event following either a main steam line break LOCA or recirculation line break LOCA, it was demonstrated that a postulated flaw will remain stable and will not propagate throughout the event (i.e., crack driving force (KIapplied) is less than the allowable Mode I, plain strain, static initiation fracture toughness (KIc/1.414))

The main steam line break is the controlling transient for evaluating the effect of reactor vessel reflood thermal shock and is based on a 1/4T flaw evaluation. The results of the analysis for the main steam line break, which shows that the maximum Klapplied for the beltline shells during a main steam line break LOCA is 105 ksiin and is less than the allowable of 141 ksiin (KIC/1.414), are shown below in Table 4.2-6. Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 10 Page 3 of 4

Table 4.2-6 Crack Stability Analysis for Beltline Shells during Main Steam Line Break Minimum vessel temperature (°F) 280

Limiting ART at 0T at 54 EFPY (°F) 83.1

T-ART (°F) 196.9

KIc/1.414 (ksiin) 141

Maximum KIapplied (ksiin) 105

Margin 1.3

The results of the analysis for the recirculation line break, which shows that the Klapplied for beltline shells during a recirculation line break LOCA is a maximum of 56 ksiin at 480 seconds and is less than the allowable of 92 ksiin (KIC/1.414), are shown below in Table 4.2-7.

Table 4.2-7 Crack Stability Analysis for Beltline Shells during Recirculation Line Break Time during transient 0 25 84 480 1200 3000 (s)

Temperature at 0.052t 550 450 450 160 120 80 (°F)

Limiting ART at 0t at 54 83.1 83.1 83.1 83.1 83.1 83.1 EFPY (°F)

T-ART (°F) 467 367 367 76.9 36.9 -3.1

KIc/1.414 (ksiin) 141 141 141 92 54 37

KIapplied at 0.052t 33 47 20 56 47 24 (ksiin)

Margin 4.2 3.0 7.0 1.6 1.1 1.5

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 10 Page 4 of 4

Beltline Nozzles

The low pressure coolant injection (LPCI) N6 nozzles do not exceed the threshold of 1.0E+17 n/cm2 prior to 54 EFPY. The water level instrument nozzle (WLIN) N12 nozzles exceed the threshold and is considered to be in the beltline. These nozzles are not bounding for the present evaluation, both because of the thermal transient conditions and because there is no significant pressure load at the time of the maximum thermal stresses. Consequently, the geometric discontinuity caused by the penetrations does not contribute a localized intensification of the crack driving force.

The RPV reflood thermal shock analysis has been projected to the end of the period of extended operation.

Disposition: 10 CFR 54.21(c)(1)(ii) The RPV reflood thermal shock analysis has been projected to the end of the period of extended operation

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 11 Page 1 of 6

1

LRA Section: Section 4.3.2, Table 4.3-3A

LRA Page Number(s): Page 4.3-12 and Page 4.3-13

Reference:

TRP-060.2-03-01

Description of Change: LRA Section 4.3.2 and Table 4.3-3A is revised to provide additional detail on the 60-year projections for the non-Class 1 systems affected by thermal and pressure cycles that are different than the cycles experienced by systems connected to ASME Section III, Class 1 piping.

PNPP LRA Section 4.3.2, Pages 4.3-12 and 4.3-13, is revised as follows:

4.3.2 Non-Class 1 Fatigue

TLAA Description

Piping designed in accordance with ASME Section III, Class 2 or 3 or ANSI B31.1 Piping Code is not required to have an explicit analysis of cumulative fatigue usage, but cyclic loading is considered in a simplified manner in the design process. These codes first require prediction of the overall number of thermal and pressure cycles expected during the 40-year lifetime of these components. Then a stress range reduction factor is determined for the predicted number of cycles. If the total number of cycles is 7,000 or less, the stress range reduction factor of 1.0 is applied, which would not reduce the allowable stress value. For higher numbers of cycles, a stress range reduction factor of less than 1.0 is applied that limits the allowable stresses applied to the piping, which reduces the likelihood of failure due to cyclic loading. These are considered to be implicit fatigue analyses since they are based upon cycles anticipated for the life of the component and are therefore, TLAAs requiring evaluation for the period of extended operation. [References 9.3 (App. H, Sec.3.1.1 PDF pg 445) and 9.34 (Section 102.3.2)]

TLAA Evaluation:

Portions of some of the non-Class 1 systems, such as residual heat removal and high pressure core spray, were designed in accordance with ASME Section III, Class 2 or 3 or ANSI B31.1 requirements, but are attached to ASME Section III, Class 1 piping and are affected by the same thermal and pressure transients as the Class 1 systems. The 60-year projections for the transient types that affect these piping systems demonstrate that the total number of cycles through the period of extended operation are limited to well below 7,000 cycles, as shown in Table 4.3-1, 60-Year Projected Cycles. Therefore, the stress range reduction factors originally selected for the components within these systems Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 11 Page 2 of 6

remain applicable and these TLAAs will remain valid through the period of extended operation.

The remaining systems, listed in Table 4.3-3A below, designed in accordance with ASME Section III, Class 2 or 3 or ANSI B31.1 requirements are affected by different thermal and pressure cycles related to their specific operations.

An operational review was performed for each system to determine the number of cycles that have occurred in the past and to project the total number of cycles that will occur through the period of extended operation. This includes cycles during unit pre-operational testing, plant operational cycles, and periodic surveillance test cycles, as applicable. The results of this review are included in Table 4.3-3A.

Control and computer room humidification Reactor plant sampling Fire protection Auxiliary steam and drains Hydrogen chemistry system Post accident sampling Div. 1 & 2 standby diesel generator exhaust, intake and crankcase Emergency DG Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 11 Page 3 of 6

Table 4.3-3A - Other Non-Class 1 Systems

System Cycle Cycles 60-year Cycle Description Projection

M29 Control and Computer Room Humidification

The control and computer room humidification system is normally operated during the winter or when the Control or Computer Room humidity drops below 40%.

Cycles (winter-2/yr 120 start up/shutdown)

Cycles 10/yr 600 (humidity-start up/shutdown)

Total cycles for 720 60 years

P35 Reactor Plant Sampling

The reactor plant sampling system primarily samples utilizing a continuous flow sample stream that is not isolated between samples. Special samples may be drawn infrequently through isolated lines and result in a thermal cycle.

Cycles 1/unit start 176 (continuous flow up-lines) shutdown cycle

Cycles (isolated 52/year 3120 sample lines) Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 11 Page 4 of 6

Total cycles for 3296 60 years

P54 Fire Protection

The fire pump diesel engine is tested periodically (monthly and 18 months) and over a sixty year period would be less than 1000 starts.

Cycles (Monthly 1/mo 720 test)

Cycle (18 mo. 1/18 mo. 40 test)

Total cycles for 760 60 years

P61 Auxiliary Steam and Drains

The auxiliary steam system is normally used only during plant shutdown to provide steam to systems such as hot water heating

Cycles (Normal 1/unit start 176 operation) up-shutdown cycle

P73 Hydrogen Chemistry System

The hydrogen water chemistry system injects hydrogen into the feedwater system. The in-scope portion of the system is the mitigation monitoring system (MMS), which analyzes the amount of noble metal remaining on the monitors sample coupons, which is representative of the amount of noble metal remaining on the internal surfaces of the reactor vessel. The MMS draws and returns sample water from the Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 11 Page 5 of 6

reactor water clean up system, which it affected by the same thermal and pressure transients as the Class 1 systems.

Cycles (Normal 1/unit start 176 Operation) up-shutdown cycle

P87 Post Accident Sampling

The sample lines of the post accident sampling system (PASS) are heat traced and therefore will not experience thermal transients during sampling operations.

Cycles (Normal 1/unit start 176 Operation) up-shutdown cycle

R48 Div. 1 & 2 Standby Diesel Generator Exhaust, Intake and Crankcase & E22 Emergency DG

The diesel generators are tested periodically (monthly, 6 months, 2 years and 10 years). Including the pre-operational and start-up testing, the number of starts of each diesel generator over a sixty year period would be less than 2000 starts.

Cycles (Est. Pre-300 Op and Start-Up Testing)

Cycles (Monthly 1/mo 720 test)

Cycles (6 mo. 1/6 mo. 120 test)

Cycles (2 yr. 1/2 yrs 30 test) Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 11 Page 6 of 6

Cycles (10 yr. 1/10 yrs. 6 test)

Cycles 5/yr 300 (Maintenance starts)

Total cycles for 1476 60 years

An operational review was performed for each system to determine the number of cycles that have occurred in the past and to project the total number of cycles that will occur through the period of extended operation. This includes cycles during unit pre-operational testing, plant operational cycles, and periodic surveillance test cycles, as applicable. For each of these systems, the review concluded that the total number of cycles, projected for 60 years, will not exceed 7,000 cycles.

The transient cycles for the non-Class 1 systems have been evaluated and indicate that 7,000 transient cycles will not be exceeded for 60 years of operation. Therefore, the non-class 1 piping stress analyses remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

Disposition: 10 CFR 54.21(c)(1)(i) The non-Class 1 piping stress analyses remain valid for the period of extended operation.

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 12 Page 1 of 4

2

LRA Section: Section 4.3.3

LRA Page Number(s): Page 4.3-14 through Page 4.3-15

References:

TRP-060.3-02 and TRP-060.3-03

Description of Change: LRA Section 4.3.3 is revised to:

Provide the specific NUREG/CR-6260, Appendix A, formula used to determine the Fen each applicable material.

Clarify that the underlined 60-Year Uen values in Table 4.3-5 are the NUREG/CR-6260 locations.

PNPP LRA Section 4.3.2, Pages 4.3-14 and 4.3-15, is revised as follows:

4.3.3 ENVIRONMENTAL FATIGUE

TLAA

Description:

NUREG-1800, Revision 2 [Reference 4.7-20], provides a recommendation for evaluating the effects of the reactor water environment on the fatigue life of ASME Section III Class 1 components that contact reactor coolant. One method acceptable to the staff for satisfying this recommendation is to assess the impact of the reactor coolant environment on a sample of critical components. These critical components should include those selected in NUREG/CR-6260 [Reference 4.7-21] that are applicable to the plant. Consideration should also be given to adding additional component locations if they are considered to be more limiting than those considered in NUREG/CR-6260.

TLAA Evaluation:

NUREG/CR-6260 identified the following component locations, which are directly relevant to PNPP, to be the most sensitive to environmental effects for newer vintage General Electric plants.

1. Reactor vessel shell and lower head

2. Reactor vessel feedwater nozzle

3. Reactor recirculation piping (including inlet and outlet nozzles)

4. Core spray line reactor vessel nozzle and associated Class 1 piping

5. Residual heat removal nozzles and associated Class 1 piping

6. Feedwater line Class 1 piping Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary)

L-24-189 Attachment 12 Page 2 of 4

Similar to the evaluation performed in NUREG/CR-6260, the locations of highest design cumulative usage factor (CUF) were evaluated for the components listed above. In addition, consideration was given to the differing environmental fatigue correction factor (Fen) associated with different materials, since some locations are composed of multiple materials. This could result in a location with a lower design CUF having a higher resultant CUF once the Fen factor is applied. Therefore, multiple locations were evaluated for components with multiple material types.

In determining the Fen for the component materials, the overall percentage of time the plant will have operated within each water chemistry regime by the end of the period of extended operation was determined based upon a review of plant chemistry records. The three chemistry regimes for PNPP are: No hydrogen water chemistry (No HWC), Hydrogen water chemistry with noble metal chemical application (HWC with NMCA), and Online noble chemistry (OLNC).

The results of the chemistry review are presented in Table 4.3-4, Water Chemistry Operational Regimes Data. From the effective percentage of time in the regime, bounding dissolved oxygen (DO) levels were determined and used as an input in determining Fen values for each material.

The cycle-based monitoring locations discussed in Sections 4.3.1.1, 4.3.1.2, and 4.3.4 were screened to eliminate locations where environmental assisted fatigue (EAF) is not applicable to the component material/environment combination or where the CUF values are very low. For those locations not eliminated by the EAF susceptibility and CUF screening, bounding values of Fen were calculated based upon the following applicable formulas provided in Appendix A of NUREG/CR-6909 [Reference 4.7-23] for the carbon and low alloy steel and the austenitic stainless steel materials.

Carbon Steel

The environmentally assisted fatigue correction factor (Fen) for carbon steel (CS) is calculated using NUREG/CR-6909, Equation A.2.

Low Alloy Steel

The environmentally assisted fatigue correction factor (Fen) for low alloy steel (LAS) is calculated using NUREG/CR-6909, Equation A.3. Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 12 Page 3 of 4

Wrought and Cast Austenitic Stainless Steels

The environmentally assisted fatigue correction factor (Fen) for wrought and cast austenitic stainless steels is calculated using NUREG/CR-6909, Equation A.9.

Nickel-Chromium-Iron (Ni-Cr-Fe) Alloys

The environmentally assisted fatigue correction factor (Fen) for Ni-Cr-Fe alloys is calculated using NUREG/CR-6909, Equation A.14.

Location-specific DO levels and maximum temperatures were used in these Fen calculations.

Using the resulting location-specific bounding Fen values and bounding CUF values, the environmental fatigue was determined using the fatigue curves specified in NUREG/CR-6909, Tables A.1 and A.2. Section A.2.1 of NUREG/CR-6909, Revision 1 [Reference 4.7-24] allows for the use of the average temperature (i.e., average of the maximum temperature for the transient and the higher of the threshold temperature for the material under consideration and the minimum temperature for the transient) for simple transients in cases of constant strain rate and linear temperature response.

Section 4.2.6 of EPRI Report Number TR-1012017 indicates that for load pairs that may be subject to dynamic loading, Fen = 1.0 for the dynamic portion of the strain for the load pair in question. This is based on the premise that the cycling due to dynamic loading occurs too quickly for environmental effects to be significant. Accordingly, transient pairs which have solely dynamic loading values or have rapid cycling strain amplitudes below the strain amplitude threshold will have no environmental fatigue multipliers applied (i.e., Fen = 1.0).

Based on the calculated environmental fatigue values, the locations requiring more detailed analysis or monitoring were identified. The results of the EAF calculations for these bounding locations are provided in Table 4.3-5, Environmental Fatigue Evaluation Summary Results. The NUREG/CR-6260 locations are the underlined 60-year Uen values in the table.

The Fatigue Monitoring Program calculates CUFs for the limiting locations and requires corrective actions if design limits are approached. This is accomplished by the use of cycle-based fatigue (CBF) monitoring where fatigue is computed from counted transients and parameters to ensure that the CUFs for the limiting components do not exceed the design limit of 1.0, including environmental effects where applicable. Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 12 Page 4 of 4

Therefore, effects of environmental fatigue on the intended functions of the RCPB components will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21(c)(1)(iii).

Disposition: 10 CFR 54.21(c)(1)(iii) The effects of environmental fatigue on the intended functions of the RCPB components will be managed for the period of extended operation by the Fatigue Monitoring Program.

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3

LRA Section: Section 4.3.3, Table 4.3-5

LRA Page Number(s): Page 4.3-17

Reference:

TRP-060.3-02

Description of Change: LRA Table 4.3-5 is revised to add an underline to the 60-Year Uen value for HPCS piping to indicate that it is a NUREG/CR-6260 location. This underline was inadvertently omitted.

Note: The 60-Year Uen value for HPCS piping has been underlined to indicate that it is a NUREG/CR-6260 location. The usual method of identifying additional information (i.e., RED, bold and underline) would obscure the change being made. Instead, the table cell for this value has been boxed in RED.

PNPP LRA Section 4.3.2, Pages 4.3-14 and 4.3-15, is revised as follows:

Table 4.3-5 Environmental Fatigue Evaluation Summary Results

60-Year Ave. 60-Year System Location Material U Fen Uen

RPV Region A Steam dryer bracket SS 0.110 4.5 0.495

RPV Region C Support skirt/lower head LAS 0.184 3.62 0.666

FW nozzle and FW nozzle SS 0.304 3.13 0.951 piping

RHR/LPCI nozzle SS 0.043 4.0 0.172

RHR/LPCI nozzle Ni-Cr-Fe 0.139 2.49 0.346

RHR/LPCI RHR/LPCI nozzle LAS 0.119 3.80 0.452 nozzle and piping RHR/LPCI nozzle CS 0.013 4.61 0.060

LPCI piping pt. 15A-J-32-TTJA CS 0.091 3.26 0.297

LPCI piping pt. 1-I-1-TTJA CS 0.060 5.08 0.305 Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 13 Page 2 of 2

HPCS/LPCS nozzle Ni-Cr-Fe 0.266 1.84 0.491

Core spray HPCS/LPCS nozzle LAS 0.045 3.44 0.155 nozzles and piping HPCS piping pt. 46-J-27-TTJA CS 0.004 8.6 0.036

HPCS/LPCS nozzle CS 0.056 2.71 0.152

Recirculation piping pt. 216 SS 0.187 4.64 0.867 Recirc nozzles RI nozzle SS 0.036 3.78 0.136 and piping (includes RHR RO nozzle LAS 0.212 3.32 0.705 suction piping) RHR suction piping pt. 544 CS 0.094 7.31 0.687 Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 14 Page 1 of 2

4

LRA Section: Section 4.3.5 and Section 4.3.6

LRA Page Number(s): Pages 4.3-19 and 4.3-20

Reference:

TRP-060.5-01

Description of Change: LRA Section 4.3.5 is revised for the discussion of the Intermediate HELB Location Determination TLAA as follows:

The title of Section 4.3.5 was revised to the Class 1 Intermediate HELB Location Determination to differentiate it from the Class 2 and 3 Intermediate HELB Location Determination TLAA, which was added in Section 4.3.6.

Section 4.3.6 was added to provide a discussion of the Class 2 and 3 Intermediate HELB Location Determination TLAA. This TLAA discussion provides a more complete evaluation of the time dependency of the Class 2 and 3 fatigue and its potential effect on the location of intermediate HELB locations.

PNPP LRA Section 4.3.5, Page 4.3-19 is revised as follows:

4.3.5 CLASS 1 INTERMEDIATE HELB LOCATION DETERMINATION

The only change to Section 4.3.5 is the title change to differentiate the Class 1 Intermediate HELB Location Determination TLAA from the Class 2 and 3 Intermediate Helb Location Determination TLAA discussed in the Section 4.3.6, added below. Therefore, the rest of the Class 1 Intermediate HELB Location Determination TLAA discussion has not been included in this attachment.

PNPP LRA Section 4.3.6, Page 4.3-20 is added as follows:

4.3.6 CLASS 2 AND 3 INTERMEDIATE HELB LOCATION DETERMINATION

TLAA

Description:

USAR Section 3.6.2.1.5 indicates that the determination of intermediate HELB locations for Class 2 and 3 piping breaks are postulated where the piping system stress is greater than 0.8 (1.2 Sh + Sa), except where break exclusion rules apply. The allowable stress range for expansion stress Sa is a function of the allowable material stress at maximum cold and hot temperatures and the total number of full temperature cycles over the expected plant lifetime. Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 14 Page 2 of 2

Since the Sa value contains a stress range reduction factor which is related to the number of cycles the system experiences during the 40-year plant life, this evaluation is a TLAA.

TLAA Evaluation:

As discussed in Section 4.3.2, Class 2 and 3 piping designed in accordance with ASME Section III, Class 2 or 3 is not required to have an explicit analysis of cumulative fatigue usage, but cyclic loading is considered in a simplified manner in the design process.

As discussed in Section 4.3.2, the total number of cycles for the Class 2 and 3 piping will not exceed 7,000 cycles during the 60-year operating term. The stress range reduction factor remains equal to 1.0 for the Class 2 and 3 piping thus the no additional Class 2 and 3 locations will have circumferential or longitudinal stresses in excess of 0.8 (1.2 Sh + Sa). Therefore, the UFSAR Section 3.6.2.1.5 break location criteria remain unchanged during the PEO and the Class 2 and 3 intermediate HELB location determination remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

Disposition: 10 CFR 54.21(c)(1)(i) The Class 2 and 3 intermediate HELB location determination will remain valid for the period of extended operation.

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5

LRA Section: Section 4.5.3

LRA Page Number(s): Page 4.5-3

Reference:

TRP-063-03 Description of Change: LRA Section 4.5-3 is revised to incorporate the containment piping penetrations bellows into the Fatigue Management Program to manage bellows fatigue for the Period of Extended Operation (PEO). Though an evaluation of the 40-year containment piping penetrations bellows fatigue indicated that the bellows would not exceed the fatigue limits at 60-years, it was determined to make the monitoring of the containment piping penetrations bellows consistent with other components for which fatigue is an aging effect and incorporate the bellows into the Fatigue Monitoring Program.

PNPP LRA Section 4.5.3, Page 4.5-3, is revised as follows:

4.5.3 CONTAINMENT PIPING PENETRATION BELLOWS

TLAA

Description:

Guard pipe assemblies associated with containment penetrations utilize bellows. The PNPP specification required these bellows to be analyzed for at least 500 cycles of normal operation plus one safe shutdown earthquake (SSE) cycle for 40 years of operation. Therefore, these fatigue analyses are identified as TLAAs requiring disposition for license renewal.

TLAA Evaluation:

PNPP has evaluated the containment piping penetrations bellows fatigue and determined, based on the 40-year CUF for the bounding penetration bellows, that the bellows fatigue usage is bounded by the fatigue usage of the penetrations. Therefore, since the penetration fatigue is not expected to exceed allowable limits at 60-years neither would the bellows fatigue.

However, since PNPPs method for managing fatigue is to track and evaluate transient cycles and to calculate CUFs to ensure that the CUFs for the limiting components do not exceed design limits, the effects of fatigue on the containment piping penetrations bellows will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21(c)(1)(iii).The fatigue analyses for these bellows determined they were capable of handling the Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 15 Page 2 of 2

movement from more normal operation or faulted cycles than were specified. The bellows are qualified for more than the 60 year projected number of startups and shutdowns, and therefore, the bellows analyses remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

Disposition: 10 CFR 54.21(c)(1)(iii)(i) The effects of fatigue on the containment piping penetrations bellows will be managed for the period of extended operation by the Fatigue Monitoring Program. The bellows fatigue analyses will remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

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6

LRA Section: Section 4.6.8

LRA Page Number(s): Page 4.6-9

Reference:

The LRA change is initiated by VistraOps staff, i.e., applicant initiated.

Description of Change: LRA Section 4.6-8 is revised to clarify that for the Engineered Safety Feature (ESF) system ductwork silicone sealant testing that PNPP is committed to monitor two samples of the ductwork/sealant combination rather than one sample.

PNPP LRA Section 4.6.8, Page 4.6-9, is revised as follows:

4.6.8 Silicone Sealant in Engineered Safety Features (ESF) HVAC Ductwork

TLAA

Description:

Silicone sealants are used in the ESF HVAC ductwork on a limited basis. PNPP performed a qualification program to demonstrate the capability of the sealant to perform its intended function for the 40 year life of the plant. This qualification program was submitted to the NRC be letter dated July 30, 1986 [Reference 4.7-28]. In addition to this qualification program, PNPP committed to perform routine monitoring of the applicable ESF ductwork and samples sample of a ductwork/sealant combination [Reference 4.7-36].

The qualification of this silicone sealant for use in the ESF ductwork is a TLAA.

TLAA Evaluation:

As discussed above, PNPP committed to a monitoring program for the applicable ESF ductwork and samples sample of the ductwork/sealant combination. The commitment for this monitoring program will continue into the period of extended operation and will be managed by the External Surfaces Monitoring of Mechanical Components AMP.

Disposition: 10 CFR 54.21(c)(1)(iii) The silicone sealant in the ESF ductwork will be managed during the period of extended operation by the External Surfaces Monitoring of Mechanical Components AMP.

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 17 Page 1 of 2

7

LRA Section: Section 4.6.11

LRA Page Number(s): Page 4.6-11

Reference:

TRP-116.11-01 Description of Change: LRA Section 4.6.11 is deleted because the identified TLAA in LRA Section 4.6.11, Allowable Stress Analysis of BOP ASME CODE Class 1, 2 and 3 Components, was found to be redundant to existing sections of the LRA addressing the same TLAAs:

Class 1 piping and components are addressed in LRA Section 4.3.1. Class 2 and 3 piping and components are addressed in LRA Section 4.3.2

PNPP LRA Section 4.6.11, Page 4.6-11, is revised as follows:

4.6.11 Allowable Stress Analysis of BOP ASME Code Class 1, 2 and 3 Components Deleted

TLAA

Description:

As discussed in USAR Section 3.9.3.1.2 [Reference 7.4.1.p], the balance of plant systems and components are identified in accordance with ASME Code Class and Safety Class. The design limits and load for the ASME Code Class 1, 2 and 3 systems are provided in USAR Tables 3.9-18 through 3.9-21a.

The stress calculations based on these limits are TLAAs.

TLAA Evaluation:

USAR Section 3.9.2.1 and 14.2.12 discuss the pre-operational dynamic testing performed to verify that system design stress limits are not exceeded.

The stress limits for BOP ASME Code Class 1, 2 and 3 components were determined based on the the applicable portions of the the ASME Code. Though the design requirements of the Code are somewhat affected by the assumed 40 year plant life time, these stress limits will remain applicable for the period of extended operation.

Additionally, the cumulative fatigue usage factors are addressed in LRA Sections 4.3.1.and 4.3.2. Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 17 Page 2 of 2

Disposition: 10 CFR 54.21(c)(1)(i) The allowable stress for the BOP ASME Code Class 1, 2 and 3 components remain valid for the period of extended operation.

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 18 Page 1 of 2

Attachment No. 18

LRA Section: Section 4.7

LRA Page Number(s): Page 4.7-2

Reference:

TRP-059-1-01

Description of Change: Reference 4.7-13 is added and the basis for this additional reference is addressed in Attachment 6 for LRA Section 4.2.1.

PNPP Section 4.7, Page 4.7-2, is revised as follows:

4.7-12 BWRVIP-108NP: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA, 2007 4.7-13 Not Used William H Bateman to Bill Eaton, Safety Evaluation of Proprietary EPRI Reports BWRVIP RAMA Fluence Methodology Manual (BWRVIP-114)," "RAMA Fluence Methodology Benchmark Manual (BWRVIP-115),"

"RAMA Fluence Methodology - Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5 (BWRVIP-117)," and "RAMA Fluence Methodology Procedures Manual (BWRVIP-121)" and "Hope Creek Flux Wire Dosimeter Activation Evaluation for Cycle 1 (TWE-PSE-001-R-001)"

(TAC No. MB9765), May 13, 2005 4.7-14 Ranganath, S., Fracture Mechanics Evaluation of a Boiling Water Reactor Vessel Following a Postulated Loss of Coolant Accident, Fifth International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, August 1979, Paper G1/5 4.7-15 General Electric Report No. NEDO-10029, An Analytical Study on Brittle Fracture of GE-BWR Vessels Subject to the Design Basis Accident, L.C. Hsu, June 1969 4.7-16 10 CFR 50.49 Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 4.7-17 NUREG-0588, Revision 1, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, July 1981 4.7-18 NRC Regulatory Guide 1.89, Revision 1, Environmental Qualification of Certain Electrical Equipment Important to Safety for Nuclear Power Plants, June 1984 4.7-19 CMAA-70-75, Crane Manufacturers Associate of America Specification 70, Specifications for Electric Overhead Traveling Cranes, Copyright 1975 4.7-20 NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Revision 2 Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 18 Page 2 of 2

4.7-21 NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, February 1995 4.7-22 NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2 4.7-23 NUREG/CR-6909, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, February 2007 4.7-24 NUREG/CR-6909, Rev. 1, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, Draft Report for Comment, March 2014. 4.7-25 EPRI Report No. TR-1012017, Guidance for Performing Environmental Fatigue Evaluations (MRP-47, Revision 1), Palo Alto, CA, September 2005 4.7-26 BWRVIP-74-A, BWR Vessel and Internals Project BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal, June 2003 4.7-27 Perry Nuclear Power Plant, Unit No. 1 - Issuance of Amendment Concerning Changes To Pressure Temperature Curves (TAC NO. MF4351)(L-14-150), June 12, 2015 (ML15141A482)

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 19 Page 1 of 1

Attachment No. 19

LRA Section: Appendix A, Section A.2.2.4

LRA Page Number(s): Page A-47 and A-48

Reference:

The LRA change is initiated by VistraOps staff, i.e., applicant initiated. Description of Change: LRA Section A.2.2.6 of Appendix A is revised to correct a typological error where the content of Section A.2.2.4 was a repeat of LRA Section A.2.2.3. PNPP Section 4.6.8, Page 4.6-9, is revised as follows:

A.2.2.4 PRESSURE TEMPERATURE (P-T) LIMITS

10 CFR 50 Appendix G [Reference 4.7-6] requires that the reactor pressure vessel be maintained within established pressure-temperature (P-T) limits, including heatup and cooldown operations. These limits specify the maximum allowable pressure as a function of reactor coolant temperature. As the reactor pressure vessel is exposed to increased neutron irradiation, its fracture toughness is reduced. The P-T limits must account for the anticipated reactor vessel fluence. The current PNPP P-T limits are valid to 32 Effective Full Power Years (EFPY) of operation. Since the P-T limits are a function of 32 EFPY fluence, associated with the 40-year licensed operating period, these P-T limits meet the criteria of 10 CFR 54.3(a) and have been identified as TLAAs requiring evaluation for 60 years.

The P-T limits are located in the Technical Specifications and are required to be updated through the 10 CFR 50.90 licensing process separate from this license renewal process in accordance with 10 CFR 54.21(c)(1)(iii).

10 CFR 50, Appendix G, defines the fracture toughness requirements for the life of the vessel. The shift in the initial RTNDT (RTNDT) is evaluated as the difference in the 30 ft-lb index temperatures from the average Charpy curves measured before and after irradiation. This increase (RTNDT) determines how much higher the vessel temperature must be raised for the material to continue to act in a ductile manner. The ART is defined as: Initial RTNDT + RTNDT + Margin. Since the RTNDT value is a function of 32 EFPY fluence in the current ART calculations associated with the 40-year licensed operating period, these ART evaluations are TLAAs requiring evaluation for 60 years.

54 EFPY ART values have been evaluated and are projected to remain within the acceptable limits through the period of extended operation. The ART analyses have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii). Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 20 Page 1 of 3

Attachment No. 20

LRA Section: Appendix B, Section B.2.14

LRA Page Number(s): Page B-48 through B-50

Reference:

The LRA change is initiated by VistraOps staff, i.e., applicant initiated.

Description of Change: LRA Section B.2.14 of Appendix B, BWR Vessel Internals Program is revised to correct a typological error.

PNPP Section B.2.14, Page B-49, is revised as follows:

B.2.14 BWR VESSEL INTERNALS PROGRAM The BWR Vessel Internals Program is an existing condition monitoring program that includes inspection and flaw evaluation in accordance with the requirements of ASME Code, Section XI and Boiling Water Reactor Vessel and Internals Project (BWRVIP) reports. The program manages the effects of cracking, loss of material and loss of fracture toughness of vessel internal components in a reactor coolant or steam environment. Reactor coolant water chemistry is controlled and monitored to maintain high water purity and reduce susceptibility to stress corrosion cracking (SCC) or intergranular stress corrosion cracking (IGSCC) as described in the Water Chemistry Program.

The BWR Vessel Internals Program incorporates the inspection and flaw evaluation (I&E) recommendations and the repair design criteria guidelines for the in-scope components as identified in the following listing:

In-Scope Component Applicable I&E Applicable Repair BWRVIP BWRVIP

Core Shroud BWRVIP-76, R-1-A BWRVIP-02-A

Core Plate BWRVIP-25, R-1-A BWRVIP-50-A

Core Spray BWRVIP-18, R-2-A BWRVIP-16-A; BWRVIP-19A

Shroud Support BWRVIP-38 BWRVIP-52-A

Jet Pump Assembly BWRVIP-41, R-4-A; BWRVIP-138, R-1-A BWRVIP-51-A Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 20 Page 2 of 3

Low Pressure Coolant Core BWRVIP-42, R-1-A BWRVIP-56-A Injection Couplings

Top Guide BWRVIP-26-A BWRVIP-50-A BWRVIP-183-A Control Rod Drive Housing BWRVIP-47-A BWRVIP-58-A

Lower Plenum Components BWRVIP-47-A BWRVIP-57 R-1

Steam Dryer BWRVIP-139, R-1-A BWRVIP-181 R-2

Access Hole Cover BWRVIP-180 R-1 BWRVIP-217

Orificed Fuel Support BWRVIP-47-A N/A

The BWR Vessel Internals Program indirectly manages the loss of fracture toughness due to neutron or thermal embrittlement of cast austenitic stainless steel (CASS) and Alloy X-750 materials. The program relies on the assessment of the susceptibility of the CASS component materials for thermal aging and neutron irradiation embrittlement performed in BWRVIP-234-A and the assessment of the X-750 component materials in BWRVIP-138-A.

NUREG-1801 Consistency

The BWR Vessel Internals Program is an existing Perry program that is consistent, with enhancement, with the 10 elements of an effective aging management program as described in NUREG-1801, Section XI.M9 BWR Vessel Internals.

Exceptions to NUREG-1801:

None.

Enhancements:

The following enhancements will be completed 6 months prior to entering the period of extended operation:

An evaluation of the 60-year fluence for the six (6) critical components identified in BWRVIP-234, Table 6-1, will be performed to verify the applicability of the BWRVIP to PNPP. In the unlikely circumstance that the 60-year fluence limits for one or more these components are exceeded, an assessment of the susceptibility of reactor vessel internal components fabricated from CASS to loss Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 20 Page 3 of 3

of fracture toughness due to thermal aging and neutron irradiation embrittlement will be performed. The required periodic inspections of CASS components determined to be susceptible to loss of fracture toughness due to thermal aging and neutron irradiation embrittlement will be determined based on this assessment. Program Elements Affected: Scope of Program (Element 1) and Acceptance Criteria (Element 6)

The BWR Vessel Internals Program implementing station procedures will be revised to incorporate BWRVIP-14, BWRVIP-59, and BWRVIP-60 as guidelines for evaluation of crack growth in stainless steels, nickel alloys, and low-alloy steels, respectively. Program Element Affected: Monitoring and Trending (Element 5).

Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 21 Page 1 of 2

Attachment No. 21

LRA Section: Appendix C

LRA Page Number(s): Page C-10

Reference:

TRP-009-01

Description of Change: LRA Appendix C is revised to clarify how the BWRVIP documents are referenced by the PNPP BWR Vessel Internals Program.

PNPP Appendix C, C-10, is revised as follows:

Note: Only the changed pages of the affected LRA table are provided

LRA Appendix C

BWRVIP-74-A (14) No ASME Section XI flawed components with evaluations in accordance with Components that have indications that have subsection IWB-3600 of Section XI to the been previously analytically evaluated in ASME Code until the end of the 40-year accordance with sub-section IWB-3600 of service period were identified for PNPP. Section Xl to the ASME Code until the end of the [Section 3.6.4] 40-year service period shall be reevaluated for the 60 year service period corresponding to the LR term.

BWRVIP-76 Rev. 1, BWR Core Shroud Inspection and Flaw Evaluation Guidelines

BWRVIP-76, Rev. 1 (4) The PNPP BWR Vessel Internals Program discussed in LRA Appendix B, Section The applicants shall reference the NRC staff B.2.14, references BWRVIP-14-A, approved TRs BWRVIP-14-A, BWRVIP-99 (when BWRVIP-99-A and BWRVIP-100-A. The approved) and BWRVIP-100-A in their RVI PNPP BWR Vessel Internals Program components' AMP. The applicants shall make a discussed in LRA Appendix B, Section statement in their LRAs that the crack growth B.2.14, cites BWRVIP-14-A. BWRVIP-rate evaluations and fracture toughness values 99-A and BWRVIP-100-A are referenced specified in these reports shall be used for in the PNPP BWR Vessel Internals cracked core shroud welds that are exposed to Program implementing documents. the neutron fluence values that are specified in these TRs. The applicants shall confirm that they Any emerging inspection guidelines will incorporate any emerging inspection developed by the BWRVIP for these welds guidelines developed by the BWRVIP for these will be evaluated and incorporated as welds. applicable. Perry Nuclear Power Plant LRA Supplement 1 (Non-Proprietary) L-24-189 Attachment 21 Page 2 of 2

BWRVIP-76, Rev. 1 (5) PNPP does not have a core shroud with tie rod repairs. LR applicants that have core shrouds with tie rod repairs shall make a statement in their AMPs associated with the RVI components that they have evaluated the implications of the Hatch Unit 1 tie rod repair cracking on their units and incorporated revised inspection guidelines, if any, developed by the BWRVIP.}}