L-24-257, License Renewal Application for the Perry Power Plant Revision 0 - Response to Requests for Confirmatory Information - (Set 2)

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License Renewal Application for the Perry Power Plant Revision 0 - Response to Requests for Confirmatory Information - (Set 2)
ML24339A066
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 12/04/2024
From: Penfield R
Vistra Operations Company
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
L-24-257
Download: ML24339A066 (1)
v • d • e


Text

L-24-257 December 4, 2024 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 Perry Nuclear Power Plant Rod L. Penfield Site Vice President 1 O Center Road Perry, Ohio 44081 10 CFR 54 License Renewal Application for the Perry Nuclear Power Plant Revision O - Response to Requests for Confirmatory Information - (Set 2)

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 O (ADAMS Accession No. M L23184A081 )
2. Nuclear Regulatory Commission issuance of Conforming License Amendment 203 to Facility Operating License NPF-58 (Enclosure 1) for the license transfer for the Perry Nuclear Power Plant (ADAMS Accession Nos. ML24057A075 and ML24057A077)
3. 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)
4. 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. ML232618019)
5. Letter L-24-189, from Rod L. Penfield to the Nuclear Regulatory Commission, dated August 7, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 1 (Non-Proprietary) (ADAMS Accession No. ML24220A270) 6555 SIERRA DRIVE IRVING, TEXAS 75039 o 214-812-4600 VISTRACORP.COM

Perry Nuclear Power Plant L-24-257 Page 2 of 3

6. Letter L-24-020, from Rod L. Penfield to the Nuclear Regulatory Commission, dated June 27, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 2 (ADAMS Accession No. ML24180A010)
7. Letter L-24-108, from Rod L. Penfield to the Nuclear Regulatory Commission, dated July 24, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 3 (ADAMS Accession No. ML24206A150)
8. Letter L-24-200, from Rod L. Penfield to the Nuclear Regulatory Commission, dated September 5, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 4 Revision 1 (ADAMS Accession No. ML24249A123)
9. Letter L-24-179, from Rod L. Penfield to the Nuclear Regulatory Commission, dated October 21, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 5 (ADAMS Accession No. ML24295A352)
10. Letter L-24-243 from Rod L. Penfield to the Nuclear Regulatory Commission, dated November 7, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 6 (ADAMS Accession No. ML24312A368)
11. Letter L-24-207, from Rod L. Penfield to the Nuclear Regulatory Commission, dated September 16, 2024, submitting the License Renewal Application for the Perry Nuclear Power Plant - Response to Request for Additional Information - Set 1 (ADAMS Accession No. ML24260A266)
12. Letter L-24-208, from Rod L. Penfield to the Nuclear Regulatory Commission, dated October 2, 2024, submitting the License Renewal Application for the Perry Nuclear Power Plant -

Response to Request for Additional Information - Set 2 (ADAMS Accession No. ML24276A083)

13. Letter L-24-209, from Rod L. Penfield to the Nuclear Regulatory Commission, dated November 19, 2024, submitting the License Renewal Application for the Perry Nuclear Power Plant - Response to Request for Additional Information - Set 3 (ADAMS Accession No. ML24324A185)
14. Letter L-24-226, from Rod L. Penfield to the Nuclear Regulatory Commission, dated October 31, 2024, submitting the License Renewal Application for the Perry Nuclear Power Plant

- Response to Requests for Confirmatory Information - Set 1 (ADAMS Accession No.

M L24305A 134)

15. NRC Email from Vaughn Thomas to Rod Penfield - dated November 4, 2024 - Perry LRA -

Requests for Confirmatory Information (Set 2) (ADAMS Accession Nos. ML24309A168)

On July 3, 2023, Energy Harbor Nuclear Corp. submitted a license renewal application (LRA) for the Facility Operating License for the Perry Nuclear Power Plant, Unit No. 1 (PNPP) (Reference 1 ).

Subsequent to the submittal of the PNPP 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 (Reference 2). The license transfer changes impacting the PNPP LRA are documented in the annual amendment required by 1 O CFR 54.21 (b ), submitted on July 3, 2024 (Reference 3).

6555 SIERRA DRIVE IRVING, TEXAS 75039 o 214-81.2-4600 VISTRACORP.COM

Perry Nuclear Power Plant L-24-257 Page 3 of 3 During the Nuclear Regulatory Commission (NRC) staff's aging management audit of the PNPP LRA (Reference 4), the PNPP Staff agreed to supplement the LRA with clarifying information which has led to several LRA supplements (References 5 through 10).

In addition, as a result of the NRC's review and audit of the PNPP LRA, the NRC Staff has submitted and the PNPP Staff responded to three sets of Requests for Additional Information (RAls) (References 11, 12 and 13) and one set of Requests for Confirmatory Information (RCls) (Reference 14).

On November 4, 2024, the NRC Staff submitted the second set of RCls (Reference 15). Attachments 1 and 2 of this letter, provide the responses to these RC ls.

The regulatory commitments identified in Appendix A, Table A.3 of the PNPP LRA are not impacted by the attached responses to the RCls. If there are any questions or if additional information is required, please contact Mr. Mark Bensi, PNPP License Renewal 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 December 4, 2024.

Sincerely,

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Rod L. Penfield /

0 Attachments:

1. PNPP Response to RCI ESEB RCl-10395-R1 - RCI Question 1
2. PNPP Response to RCI ESEB RCl-10395-R1 - RCI Question 2 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 6555 SIERRA DRIVE IRVING. TEXAS 75039 o 21.4-812-4600 VISTRACORP.COM

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 1 Page 1 of 5 ESEB RCI-10395-R1 Request for Confirmation - Question 1 Regulatory Basis Part 54 of Title 10 of the Code of Federal Regulations, Requirements for renewal of operating licenses for nuclear power plants, is designed to elicit application information that will enable the NRC staff to perform an adequate safety review and the Commission to make the necessary findings. Reliability of application information is important and advanced by requirements that license applications be submitted in writing under oath or affirmation and that information provided to the NRC by a license renewal applicant or required to be maintained by NRC regulations be complete and accurate in all material respects. Information that must be submitted in writing under oath or affirmation includes the technical information required under 10 CFR 54.21(a) related to assessment of the aging effects on structures, systems, and components subject to an aging management review. Thus, both the general submission requirements for license renewal applications and the specific technical application information requirements require that submission of information material to NRCs safety findings (see 10 CFR 54.29 standards for issuance of a renewed license) be submitted by an applicant as part of the application.

Background

LRA Tables 3.5.2-1 through 3.5.2-4, as modified by LRA Supplement 3 (ML24206A150),

provide the results of component, material, environment, and aging effect combinations requiring aging management and associated aging management programs for containments, structures, and component supports. The staff noted that aging mechanisms for non-GALL items citing plant specific notes (i.e., G, H, and J) in LRA Tables 3.5.2-1 through 3.5.2-4 are not provided. The staff reviewed LRPY-CAMP-001, Structural Material/Environment/Aging Effect Bases Report - Rev.6 on the Portal and EPRI report 1015078, Plant Support Engineering:

Aging Effects for Structures and Structural Components (Structural Tools), which describe aging effects/aging mechanisms for the structural aging management review items (AMRs). The staff summarized aging mechanisms for the non-GALL items and noted that their aging effects are managed by the Structures Monitoring program, in the following table.

Component Type Material Environment Aging Effect Requiring Management Aging Management Programs Notes Aging Mechanism Structural bolting 2 (Table 3.5.2-4, item 344)

High strength steel Treated water (Ext)

Cracking Structures Monitoring G, 512 Stress Corrosion Cracking (SCC)

Structural bolting 2 (Table 3.5.2-4, item 345)

High strength steel Treated water (Ext)

Loss of Material Structures Monitoring G, 519 General corrosion/pitting corrosion Structural bolting (Table 3.5.2-4, item 333)

High strength steel Raw water (Ext)

Cracking Structures Monitoring G, 535 SCC

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 1 Page 2 of 5 Component Type Material Environment Aging Effect Requiring Management Aging Management Programs Notes Aging Mechanism Structural bolting (Table 3.5.2-4, item 334)

High strength steel Raw water (Ext)

Loss of Material Structures Monitoring G, 536 MIC, general corrosion, galvanic corrosion Culvert, major stream (Table 3.5.2-3, item 15)

Galvanized steel Raw water (Int)

Loss of Material Structures Monitoring G, 539 General corrosion Culvert, major stream (Table 3.5.2-3, item 16)

Galvanized steel Soil (Ext)

Loss of Material Structures Monitoring G, 539 General corrosion/galvanic corrosion/MIC Scupper cover (Roof)

(Table 3.5.2-2, item 193)

Stainless steel Air - outdoor (Ext)

Cracking Structures Monitoring H, 501 SCC Scuppers (Roof)

(Table 3.5.2-2, item 199)

Stainless steel Air - outdoor (Ext)

Cracking Structures Monitoring H, 501 SCC Conduit caps (Table 3.5.2-4, item 77)

Polymer Air - indoor, uncontrolled (Ext)

Loss of Strength Structures Monitoring H, 518 Loss of mechanical rigidity due to high temperature, UV, ozone or ionizing radiation exposure.

Anchorage/embedments (Table 3.5.2-4, item 19)

Steel Raw water (Ext)

Cracking Structures Monitoring H, 519 Hydrogen damage Structural bolting (Table 3.5.2-4, item 341)

Steel Raw water (Ext)

Cracking Structures Monitoring H, 519 SCC Penetration Sealant Flood (Table 3.5.2-4, item 267)

Elastomer Raw water (Ext)

Cracking Structures Monitoring H, 524 Delamination /

Shrinkage Roof membrane1 (Table 3.5.2-4, item 285)

Elastomer Raw water (Ext)

Cracking Structures Monitoring H, 524 Delamination /

Shrinkage Waterproofing Membranes (Table 3.5.2-4, item 354)

Elastomer Soil (Ext)

Cracking Structures Monitoring H, 524 Delamination /

Shrinkage Waterproofing Membranes1 (Table 3.5.2-4, item 355)

Elastomer Raw water (Ext)

Cracking Structures Monitoring H, 524 Delamination /

Shrinkage Sliding support (Table 3.5.2-4, item 315)

Lubrite/

Fluorogold Air - indoor, uncontrolled (Ext)

Chang in Material Properties Structures Monitoring H, 525 Irradiation Waterproofing Membranes (Table 3.5.2-4, item 353)

Elastomer Concrete (Int)

Cracking Structures Monitoring H, 527 Delamination /

Shrinkage Storm Drain (Table 3.5.2-4, item 322)

Steel Raw water (Int)

Flow Blockage Structures Monitoring H, 530 Debris accumulation Storm Drain1 (Table 3.5.2-4, item 325)

Polymer Raw water (Int)

Flow Blockage Structures Monitoring H, 530 Debris accumulation Culvert, major stream (Table 3.5.2-3, item 14)

Galvanized steel Raw water (Int)

Flow Blockage Structures Monitoring H, 530 Debris accumulation Storm Drain3 (Table 3.5.2-4, item 326)

Concrete Raw water (Int)

Flow Blockage Structures Monitoring H, 530 Debris accumulation

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 1 Page 3 of 5 Component Type Material Environment Aging Effect Requiring Management Aging Management Programs Notes Aging Mechanism Storm Drain3 (Table 3.5.2-4, item 328)

Concrete Soil (Ext)

Loss of Material Structures Monitoring H, 531 Corrosion of embedded steel reinforcing, reaction with aggregates Storm Drain3 (Table 3.5.2-4, item 327)

Concrete Raw Water (Int)

Loss of Material Structures Monitoring H, 531 Corrosion of embedded steel reinforcing, reaction with aggregates Roof scupper (Table 3.5.2-4, item 286)

Aluminum Air - outdoor (Ext)

Cracking Structures Monitoring H, 534 SCC Upper containment pool gates seals1 (Table 3.5.2-1, item 173)

Elastomer Treated water (Ext)

Cracking Structures Monitoring H, 537 Delamination /

Shrinkage Shielding (Table 3.5.2-4, item 314)

Unimpregna ted fiberglass fabric; Fiberglass fabric impregnated with elastomer Air - indoor, uncontrolled (Ext)

Change in Material Properties and Cracking Structures Monitoring J, 523 Ionizing radiation For anchorage/embedments (LRA Table 3.5.2-4, AMR item 19), the staff noted in LRPY-CAMP-001 (page 56) that hydrogen damage is a potential aging mechanism, but it was screened out as hydrogen damage was not an aging mechanism of concern for most structural steel materials of low yield strength.

For sliding support (Table 3.5.2-4, AMR item 315), the staff noted in LRPY-CAMP-001 (page

40) that irradiation is a potential aging mechanism for Lubrite/ Fluorogold but the staff further noted that this aging effect was not applicable to Lubrite.
1. Confirm the aging mechanisms identified in the last column of the table above for each Table 2 line item.
2. Confirm that hydrogen damage is not an applicable aging mechanism for this material/environment combination of AMR item 19 in LRA Table 3.5.2-4 and no aging effects/aging management program are needed for anchorage/embedments in this line item.
3. Confirm that irradiation is not an applicable aging mechanism for Lubrite of AMR item 315 in LRA Table 3.5.2-4 and no aging effects/aging management program are needed for Lubrite sliding support in this line item.

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 1 Page 4 of 5 PNPP Response Request 1: Confirm the aging mechanisms identified in the last column of the table above for each Table 2 line item.

A review of the above table, as modified by PNPP LRA Supplement 3 (Vistra Letter L-24-108),

was performed, comparing each line item to the corresponding LRA Table 2 row, and then referring to the PNPP AMR source document for details of the aging mechanisms assigned.

Based on the review, all of the aging mechanisms identified in the above table were confirmed, except for the following identified differences:

1. Table 3.5.2-4, Row 77, Conduit caps, polymer, Air - indoor, uncontrolled (Ext), loss of strength: In the PNPP staffs response to the NRCs third set of RAIs (Vistra Letter L 209), and specifically the response to ESEB RAI-10328-R1, Question 1, PNPP stated that there are no aging effects, however, the aging effect loss of strength was conservatively applied to ensure continued monitoring of this component type. See revised note in the RAI response for a full explanation.
2. Table 3.5.2-4, Row 19, Anchorage embedment, steel, Raw water (Ext), cracking: The aging effect cracking is not due to hydrogen damage as it is not an applicable aging mechanism. Instead, the aging effect cracking is due to the aging mechanism of stress corrosion cracking (SCC). See response to Request 2 below.
3. Table 3.5.2-4, Row 267, Penetration seals (flood), elastomer, Raw water (Ext), cracking:

The applicable aging effect cracking is not due to thermal exposure as it is not an applicable aging mechanism. Instead, the aging effect cracking is due to the aging mechanism of delamination/shrinkage.

4. Table 3.5.2-4, Row 315, Sliding support, Lubrite/Flurogold, Air - indoor, uncontrolled (Ext), Change in material properties: Irradiation is not an applicable aging mechanism for Lubrite (see response to Request 3 below). However, Fluorogold material is susceptible to the aging effect of change in material properties due to irradiation.
5. Table 3.5.2-4, Row 328, Storm Drain3, concrete, Soil (Ext), loss of material: For concrete in soil, neither the aging mechanisms of corrosion of embedded steel reinforcing, nor reaction with aggregates were assigned. Instead, the aging mechanism of aggressive chemicals is cited.
6. Table 3.5.2-4, Row 327, Storm Drain3, concrete, Raw water (Int), loss of material: For concrete in raw water, neither the aging mechanisms of corrosion of embedded steel reinforcing, nor reaction with aggregates were assigned. Instead, the aging mechanism of freeze-thaw is cited. In the PNPP staffs response to the NRCs third set of RAIs (Vistra Letter L-24-209), and specifically RAI ESEB RAI-10327-R1, Question 3, for the same component Storm Drain 3 (concrete) in a raw water environment, the aging effect of loss of material due to corrosion of embedded steel reinforcing was proposed to be added to LRA Table 3.5.2-4.

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 1 Page 5 of 5 Request 2: Confirm that hydrogen damage is not an applicable aging mechanism for this material/environment combination of AMR item 19 in LRA Table 3.5.2-4 and no aging effects/aging management program are needed for anchorage/embedments in this line item.

Table 3.5.2-4, Row 19, Component Type Anchorage/embedments, material Steel, environment Raw water (Ext) identifies the aging effect cracking being managed. Plant specific Note 519 is assigned to this row. The note states: Structural Monitoring Program will detect this aging effect. Structural Tools Table 4-3 provides the basis for cracking in steel bolting and Table 3-3 provides the basis for loss of material. Structural tools Table 4-3 does not identify cracking due to hydrogen damage as an aging mechanism. Structural tools Table 3-3 indicates for carbon and low alloy steel in a raw water environment: Hydrogen damage is not a mechanism of concern for structural steel due to the low yield strength of carbon and low-alloy steel. The PNPP Structural Material / Environment / Aging Effect Bases Report reflects this conclusion.

Thus, it is confirmed that hydrogen damage is not an applicable aging mechanism for this material/environment combination. However, since stress corrosion cracking is an applicable aging mechanism for steel in a raw water environment, the structures monitoring program is an applicable aging management program for this row.

Request 3: Confirm that irradiation is not an applicable aging mechanism for Lubrite of AMR item 315 in LRA Table 3.5.2-4 and no aging effects/aging management program are needed for Lubrite sliding support in this line item.

LRA Table 3.5.2-4, Row 315 identifies the aging effect Change in material properties. Plant specific Note 525 assigned to that row states: Aging effect identified per Structural Tools, Table 9-2. For the material Lubrite: that table states: Lubrite lubricants used in nuclear applications are designed for the environments to which they are exposed. They are designed with the ability to carry extremely heavy dynamic and static loads with a low coefficient of friction, to operate dry, or wet in high or low temperature conditions, withstand high intensities of radiation, and are not susceptible to corrosion. An industry experience search did not find any Lubrite degradation that could lead to the loss of intended function. Therefore, humidity, high temperature, and radiation are not significant in the aging of Lubrite.

Thus, it is concluded that irradiation is not an applicable aging mechanism for Lubrite.

There are no LRA changes are associated with this RCI response.

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 2 Page 1 of 3 ESEB RCI-10395-R1 Request for Confirmation - Question 2 Regulatory Basis Part 54 of Title 10 of the Code of Federal Regulations, Requirements for renewal of operating licenses for nuclear power plants, is designed to elicit application information that will enable the NRC staff to perform an adequate safety review and the Commission to make the necessary findings. Reliability of application information is important and advanced by requirements that license applications be submitted in writing under oath or affirmation and that information provided to the NRC by a license renewal applicant or required to be maintained by NRC regulations be complete and accurate in all material respects. Information that must be submitted in writing under oath or affirmation includes the technical information required under 10 CFR 54.21(a) related to assessment of the aging effects on structures, systems, and components subject to an aging management review. Thus, both the general submission requirements for license renewal applications and the specific technical application information requirements require that submission of information material to NRCs safety findings (see 10 CFR 54.29 standards for issuance of a renewed license) be submitted by an applicant as part of the application.

=

Background===

SRP-LR Table 1 AMR item 3.5.1-57, states that constant and variable load spring hangers; guides; stops components for loss of mechanical function due to corrosion, distortion, dirt, overload, fatigue due to vibratory and cyclic thermal loads aging effects to be managed by GALL Report AMP XI.S3, ASME Section XI, Subsection IWF. The LRA states that the applicable ASME Code for the current (fourth) 10-year inspection interval for PNPP, which commenced May 18, 2019, and expires on May 17, 2029, is ASME XI, 2013 Edition, as modified by 10 CFR 50.55a or relief granted in accordance with 10 CFR 50.55a.

ASME Section XI, 2013 Code Edition provides requirements in:

IWA-2213, for VT-3 Examination. It states, VT-3 includes examination for conditions that could affect operability or functional adequacy of constant load and spring-type supports.

IWF-2500, for Examination Requirements. It states, The following shall be examined in accordance with Table IWF-2500-1 (F-A): (d) clearances of guides and stops, alignment of supports, and assembly of support items; (e) hot or cold settings of spring supports and constant load supports.

IWF-3410, for Acceptance Standards - Component Support Structural Integrity. It states, "(a) Component support conditions which are unacceptable for continued service shall include the following: (1) deformations or structural degradations of fasteners, springs, clamps, or other support items; (4) improper hot or cold settings of spring supports and constant load supports; (6) improper clearances of guides and stops.

Mandatory Appendix VI Supplements, for 6.0 Visual Examination of Components Supports includes: (c) Hangers; (d) Variable spring type supports; (e) Restraints; (g)

Guides and stops; (h) Vibration control and sway braces.

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 2 Page 2 of 3 LRA Section B.2.5, ASME Section XI, Subsection IWF Program discusses Operating Experience related to various types of springs (e.g., variable, hangers, cans). AMP Basis Document discusses in program elements: (a) Parameters Monitored/Inspected, VT-3 Examinations of spring type supports for conditions such aging effects as debris, corrosion, wear that could affect operability or functional adequacy of spring type supports; (b)

Acceptance Criteria, conformance to ASME Section XI, Subsection Code requirements; and (c) Operating Experience OE discussed in LRA Section B.2.5. By letter dated July 24, 2024, Supplement 3 (ML24206A150) PNPP revised the LRA Table 1 AMR Item 3.5.1-57 to state: Not Applicable - Loss of mechanical function in sliding supports is managed by Structures Monitoring Program.

Request Confirm that the Supplement 3 revision to Table 1 AMR Item 3.5.1-57 as not applicable is an oversight and that constant and variable load spring hangers; guides; stops will be age managed consistent with GALL-LR by ASME Section XI, Subsection IWF Program AMP as originally reported in PNPP LRA Table 3.5.1 Summary of Aging Management Evaluations for Containments, Structures and Component Supports, Item Number 3.5.1-57.

PNPP Response Confirm that the Supplement 3 revision to Table 1 AMR Item 3.5.1-57 as not applicable is an oversight and that constant and variable load spring hangers; guides; stops will be age managed consistent with GALL-LR by ASME Section XI, Subsection IWF Program AMP as originally reported in PNPP LRA Table 3.5.1 Summary of Aging Management Evaluations for Containments, Structures and Component Supports, Item Number 3.5.1-

57.

PNPP confirms that it was an oversight that LRA Table 1, AMR Item 3.5.1-57 discussion text was changed to not applicable. During the development of Supplement 3 to the LRA (Vistra Letter L-24-108), a transposition error occurred that resulted in inadvertently making the same change to the discussion text for LRA Item 3.5.1-57 as 3.5.1-75, when only Item 3.5.1-75 was intended to be updated.

Upon further review of the SRP-LR Table 3.5-1 for Item 3.5.1-57, the table indicates that Item 3.5.1-57 applies to LR-GALL Rows III.B1.1.T-28 and III.B1.2.T-28. Components corresponding to those rows are not currently reflected in PNPP LRA Table 3.5.2-4. (Note that GALL Row III.B1.3.T-28 does not apply to PNPP, as there are no ASME Class MC supports at PNPP).

Perry Nuclear Power Plant Response to RCI ESEB RCI-10395-R1 L-24-257 Attachment 2 Page 3 of 3 Based on the above, PNPP intends to make the following changes to the LRA.

1. Discussion text to LRA Table 1, Item 3.5.1-57 will be restored to the original LRA content, as follows:

From:

Not Applicable - Loss of mechanical function in sliding supports is managed by Structures Monitoring Program.

To:

Consistent with NUREG-1801. The ASME Section XI, Subsection IWF program will manage aging of ASME Class 1, 2 and 3 supports.

2. LRA Table 3.5.2-4 will have the following new rows added:

Component Type: Component and piping supports Intended Function: SNS, SRE, SSR Material: Steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of mechanical function AMP: ASME Section XI, Subsection IWF NUREG-1801 Item: III.B1.1.T-28 Table 1 Item: 3.5.1-57 Note: A Component Type: Component and piping supports Intended Function: SNS, SRE, SSR Material: Steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of mechanical function AMP: ASME Section XI, Subsection IWF NUREG-1801 Item: III.B1.2.T-28 Table 1 Item: 3.5.1-57 Note: A Component Type: Component and piping supports1 Intended Function: SNS, SRE, SSR Material: Galvanized Steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of mechanical function AMP: ASME Section XI, Subsection IWF NUREG-1801 Item: III.B1.2.T-28 Table 1 Item: 3.5.1-57 Note: A LRA changes associated with this RCI response will be provided in a future LRA supplement.

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