Text

License Renewal Application for the Perry Nuclear Power Plant - Responses to Request for Additional Information - Round 1 (Set 3)
	ML24324A185
Person / Time
Site:	Perry
Issue date:	11/19/2024
From:	Penfield R; Energy Harbor Nuclear Corp
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	L-24-209
	Download: ML24324A185 (1)
	v • d • e

L-24-209 November 19, 2024 A TIN: 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 10 Center Road Perry, Ohio 44081 10 CFR 54 License Renewal Application for the Perry Nuclear Power Plant - Responses to Request for Additional Information - Round 1 (Set 3)

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. ML23261B019)

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-24-209 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-24-226, 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)

14. NRC Email from Vaughn Thomas to Rod Penfield - dated October 2, 2024 - Perry LRA -

Requests for Additional Information - Set 3 (ADAMS Accession Nos. ML24276A128 and ML24276A129)

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 10 CFR 54.21(b), submitted on July 3, 2024 (Reference 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).

6555 SIERRA DRIVE JRVING. TEXAS 75039 o 214-812-4600 VISTRACORP.COM

Perry Nuclear Power Plant L-24-209 Page 3 of 3 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 two sets of Requests for Additional Information (RAls) (References 11 and

12) and one set of Requests for Confirmatory Information (RCls) (Reference 13). Subsequently, on October 2, 2024 the NRC Staff submitted a third set of RAls (Reference 14).

Attachments 1 to 20 of this letter provide the responses to address the third set of RAls submitted by the NRC Staff on October 2, 2024. For ease of reference, an index listing the RAI responses is provided.

The regulatory commitments identified in Appendix A, Table A.3 of the PNPP LRA are not impacted by the attached responses to the RAls. 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 November 19, 2024.

Sincerely, Attachments:

PNPP Responses to LRA NRC RAls Set 3 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 214-812-4600 VISTRACORP.COM

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachments Index Page 1 of 2 PNPP LRA Set 3 RAI Responses Attachments Index Attachment No.

Associated RAI Applicable Request 1

ESEB RAI-10308-R1 Question 1 - Requests 1, 2 & 3 2

ESEB RAI-10308-R1 Question 2 - Requests 1, 2 and 3 3

ESEB RAI-10327-R1 Question 1 - TRP 46 AMP Requests 1, 2, 3, 4 and 5 4

ESEB RAI-10327-R1 Question 2 - TRP 46 AMP Requests 1, 2, 3, 4, 5 and 6 5

ESEB RAI-10327-R1 Question 3 - TRP 46 AMP Requests 1 and 2 6

ESEB RAI-10327-R1 Question 4 - TRP 46 AMP Request 7

ESEB RAI-10327-R1 Question 5 - TRP 46 AMP Requests 1, 2, 3 and 4 8

ESEB RAI-10328-R1 Question 1 - TRP 46 AMR Request 9

ESEB RAI-10328-R1 Question 2 - TRP 46 AMR Request 1 and 2 10 ESEB RAI-10328-R1 Question 3 - TRP 46 AMR Request 11 NCSG RAI-10332-R1 Question 1 Coolant Heat Exchanger Tube Bundle Replacement Frequency Requests 1, 2 and 3 12 NCSG RAI-10332-R1 Question 2 Fire Water System Program Enhancements and Exceptions Request

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachments Index Page 2 of 2 Attachment No.

Associated RAI Applicable Request 13 NCSG RAI-10337-R1 Question 1 (Drywell Mechanical Penetrations - Aging Management Programs)

Request 14 NCSG RAI-10337-R1 Question 2 (Pyrocrete)

Requests 1 and 2 15 NCSG RAI-10337-R1 Question 3 (Fiberglass/Alumina Silicate/Calcium Silicate/Mineral Fiber)

Requests 1, 2, 3, 4 and 5 16 NCSG RAI-10337-R1 Question 4 (Unimpregnated Fiberglass Fabric; Fiberglass Fabric Impregnated With Elastomer)

Requests 1 and 2 17 NCSG RAI-10337-R1 Question 5 (Gypsum Board/Drywall)

Request 18 NCSG RAI-10337-R1 Question 6 (Loss of Sealing)

Request 19 NCSG RAI-10339-R1 Question 1 - RAI B.2.45-1 Requests 1, 2 and 3 20 NCSG-RAI-10339-R1 Question 2 - RAI A.1.45-1 Request Note: Although proposed updates are addressed in several RAI responses, the final LRA updates will be submitted in a future PNPP LRA supplement.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 1 Page 1 of 3 ESEB EAI-10308-R1 Question 1 Regulatory Basis Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. As described in SRP-LR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL-LR Report when evaluation of the matter in the GALL-LR Report applies to the plant.

Background

SRP-LR Item 3.5-1, 019 addresses the aging effects of cracking due to expansion from reaction with aggregates in accessible areas of containment concrete including basemat and concrete fill-in annulus. SRP-LR lists AMP XI.S2, ASME Section XI, Subsection IWL to manage the aging effects.

LRA Section 3.5.2.2.1.8 states, for PNPP containment concrete: [a]ccessible concrete surfaces of the containment fill in annulus are monitored for cracking due to expansion from reaction with aggregates by the ASME Section XI, Subsection IWL program and are addressed under Item Number 3.5.1-19. Its associate LRA Item Number 3.5.1-12 also indicates that the accessible concrete of PNPP containment is managed by ASME Section XI, Subsection IWL for cracking due to expansion from reaction with aggregates, and its condition is used as an indicator of the condition of the inaccessible components.

LRA Item Number 3.5.1-19, as modified by Supplement 3 (ML24206A150), states that this item is not applicable because PNPP containment is a free-standing SCV, and its concrete foundation is integral to the reactor building basemat. The reactor building basemat is not considered to be accessible. The top surface of the annulus concrete is accessible. Aging of annulus concrete will be managed by the ASME Section XI, Subsection IWL. LRA Table 3.5.2-1 does not provide any items related to Item Number 3.5.1-19.

Issue It appears to the staff that the not applicable claim of LRA Item Number 3.5.1-19 is inconsistent with the guidance in SRP Item 3.5-1, 019. SRP Item 3.5-1, 019 provide guidance to staff to review the aging effects of cracking due to expansion from reaction with aggregates on components such as accessible areas of concrete fill-in annulus that require aging management. For LRA Item Number 3.5.1-19, the applicant stated that the top surface of the annulus concrete is accessible, while at the same time claiming that Item 3.5.1-19 is not applicable.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 1 Page 2 of 3 LRA Item Number 3.5.1-19 also contradicts LRA Section 3.5.2.2.1.8 and LRA Item Number 3.5.1-12. In LRA Section 3.5.2.2.1.8 and LRA Item Number 3.5.1-12, the applicant states that aging effects on accessible concrete surfaces of the containment fill in annulus are addressed under Item Number 3.5.1-19, whereas in LRA Item Number 3.5.1-19, the applicant claims Item 3.5.1-19 is not applicable.

Request

1. Provide justification on non-applicability claim for LRA Item Number 3.5.1-19.

2. If LRA Item Number 3.5.1-19 is deemed applicable, provide Table 2 item(s) in Table 3.5.2-1 accordingly.

3. Update the LRA accordingly based on the responses.

PNPP Response

1. Provide justification on non-applicability claim for LRA Item Number 3.5.1-19.

PNPP acknowledges that the non-applicability for LRA Item Number 3.5.1-19 is incorrect. The top surface of containment annulus concrete is monitored for cracking due to expansion from reaction with aggregates by the ASME Section XI, Subsection IWL program and is addressed under Item Number 3.5.1-19.

Therefore, LRA Table 3.5.2-1 will be revised to add a new row as follows:

LRA Table 3.5.2-1:

Component Type: Beams, columns, floor slabs and interior walls (Annulus Concrete)

Intended Function: SSR Material: Concrete Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking AMP: ASME Section XI, Subsection IWL NUREG-1801 Item: II.B3.1.CP-66 Table 1 Item: 3.5.1-19 Note: A Additionally, LRA Table 3.5.1, Item 3.5.1-19, Discussion column, will be revised to state:

Consistent with NUREG-1801 - PNPP containment is a free-standing SCV, and its concrete foundation is integral to the reactor building basemat. The top surface of the annulus concrete is accessible. This aging effect of annulus concrete will be managed by the ASME Section XI, Subsection IWL.

2. If LRA Item Number 3.5.1-19 is deemed applicable, provide Table 2 item(s) in Table 3.5.2-1 accordingly.

LRA Item 3.5.1-19 is deemed applicable. A new row for this aging effect has been provided in Item 1 above.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 1 Page 3 of 3

3. Update the LRA accordingly based on the responses.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 2 Page 1 of 6 ESEB EAI-10308-R1 Question 2 Regulatory Basis Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. As described in SRP-LR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL-LR Report when evaluation of the matter in the GALL-LR Report applies to the plant.

Background

SRP-LR AMR Item 3.5-1,049 addresses the aging effects of loss of material (spalling, scaling) and cracking due to freeze-thaw in inaccessible concrete areas of Groups 6 Structures (Water Control Structures) including exterior above-and below-grade; foundation; interior slab. SRP-LR requires further evaluation for plants located in moderate to severe weathering conditions (weathering index >100 day-inch/yr). NRCs interim staff guidance of Updated Aging Management Criteria for Structures Portions (SLR-ISG-2021-03-STRUCTURES, ML20181A381) lists plant-specific aging management program or AMP XI.S6, Structures Monitoring, enhanced as necessary, as aging management program for Item 3.5-1,049. SRP-LR Subsection 3.5.3.2.2.3, Aging Management of Inaccessible Areas for Group 6 Structures, Item 1 states that a plant-specific program is not required if documented evidence confirms that where the existing concrete had air content of 3% to 8% and subsequent inspection of accessible areas did not exhibit degradation related to freeze-thaw. LRA Section 2.4.3, Water Control Structures, describes the structures and structural components within the scope of license renewal. LRA Table 2.4.3-1 lists component types subject to aging management review.

PNPP is located in a severe weathering region per Figure 1 of American Society for Testing of Materials (ASTM) C33, Location of Weathering Regions. LRA Item Number 3.5.1-49, as modified by Supplement 3 (ML24206A150), states that this item is not applicable because:

[t]he below grade inaccessible concrete areas of PNPP Group 6 structures were constructed in a manner that minimizes the potential for any freeze-thaw aging effects. The loss of material (spalling, scaling) and cracking due to freeze-thaw are not aging effects requiring management for PNPP Groups 6 structures. The air content percentages for concrete are less than 8%. The absence of this concrete aging effects is confirmed by the Structures Monitoring Program and also based on 35 years of operating experience. The foundation levels of all groups of structures are well below the frost line which would preclude this aging effect. Therefore, loss of material (spalling, scaling) and cracking due to freeze-thaw are not aging effects requiring management for PNPP Groups 6 structures. LRA Table 3.5.2-3, Water Control Structures, does not provide any items related to Item Number 3.5.1-49.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 2 Page 2 of 6 Issue Cracking, spalling and disintegration of concrete due to freeze-thaw cycling are concrete aging effects requiring management in the environment with moderate to severe weathering conditions. Although, as stated in SRP-LR, a plant-specific program is not required if the existing concrete was constructed with air content of 3% to 8% and subsequent inspection of accessible areas shows no freeze-thaw related degradation, this does not mean that aging effects are not applicable. The potential aging effects of loss of material (spalling, scaling) and cracking due to freeze-thaw on the applicable components (i.e., Group 6 concrete inaccessible areas: exterior above-and below-grade; foundation; interior slab) still need to be adequately managed during the period of extended operation through an appropriate aging management program (e.g.,

Structural Monitoring Program).

All relevant components required by SRP-LR Item 3.5-1, 049 should be included into the aging management review. SRP-LR Item 3.5-1, 049 requires managing the aging effects in inaccessible concrete areas of Groups 6 structures including exterior above-and below-grade; foundation; interior slab, while the applicant's statement of non-applicability appears to apply only to the concrete inaccessible areas well below the frost line, seemingly excluding other inaccessible concrete areas where aging may occur from the scope of its aging management review. For example, in addition to the foundation, the emergency service water pumphouse as shown in PNPP USFAR Figure 9A-34 (DWG. E-023-0034-00000) contains exterior above-ground and below-ground concrete walls that may be inaccessible but require aging management review.

Request

1. With reference to the structures and structural components within the scope of license renewal as described in LRA Section 2.4.3, clarify whether all applicable inaccessible concrete components required by SRP-LR Item 3.5-1, 049 (i.e. external above-and below-grade; foundation; internal slab) are included in the aging management review and, accordingly, reevaluate whether the statement of non-applicability for LRA Item No. 3.5.1-49 is appropriate.

2. If LRA Item Number 3.5.1-49 is deemed applicable, identify an AMP to manage the aging effects, explain how the designated AMP will adequately manage freeze-thaw related aging effects in the inaccessible area of Group 6 concrete if they occur during the period of extended operation, and provide associate Table 2 item(s) in Table 3.5.2-3, Water Control Structures, accordingly. If LRA Item Number 3.5.1-49 is considered not applicable, provide a rationale for how inaccessible concrete areas that may be at risk for freeze-thaw cycles (e.g., portions of underground concrete walls located above or around the frost line) are excluded from aging management review.

3. Update LRA accordingly based on the responses.

PNPP Response

1. With reference to the structures and structural components within the scope of license renewal as described in LRA Section 2.4.3, clarify whether all applicable inaccessible concrete components required by SRP-LR Item 3.5-1, 049 (i.e. external above-and below-grade; foundation; internal slab) are included in the aging

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 2 Page 3 of 6 management review and, accordingly, reevaluate whether the statement of non-applicability for LRA Item No. 3.5.1-49 is appropriate.

PNPP confirms that all structures and structural components within the scope of license renewal as described in LRA Section 2.4.3 that are components included in the description of SRP-LR 3.5.1-49 (exterior above-and below-grade; foundation; interior slab) categorized as inaccessible are included in the aging management review detailed in LRA Table 3.5.2-3.

For locations that are inaccessible, LRA Section B.2.43 (Structures Monitoring Program) has an existing enhancement that states as follows:

The program will be enhanced to require (a) evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas and (b) examination of representative samples of the exposed portions of the below grade concrete, when excavated for any reason. If normally inaccessible areas become accessible due to planned activities, an inspection of these areas shall be conducted.

Based on the justification provided in LRA Sections 3.5.2.2.2.3.1 and 3.5.2.2.1.7, Table 1 Item 3.5.1-49 was considered as Not Applicable. However, PNPP agrees with NRC Staffs recommendation and will continue to include this aging effect as applicable and will use Structures Monitoring Program to manage this aging effect. LRA Table 1 Item 3.5.1-49 will be noted as Consistent with NUREG-1801 and considered applicable. Specifically, the following inaccessible Component Types in LRA Table 3.5.2-3 will include this aging effect to be managed:

1. Exterior walls below grade (Environment: Soil (ext))

2. Exterior walls above grade (Environment: Air - outdoor (Ext))

3. Floor slab (Environment: Raw water (Int))

Note: See the response to Request 2 below for details of the changes to Table 3.5.2-3 LRA Table 3.5.1, Item 3.5.1-49 discussion text will be revised as follows:

Consistent with NUREG-1801 with the following clarification - The air content percentages for concrete are less than 8%. The absence of this concrete aging effects is confirmed by the Structures Monitoring Program and also based on 35 years of operating experience. The foundation levels of all groups of structures are well below the frost line which would preclude this aging effect. The below grade inaccessible concrete areas of PNPP Group 6 structures were constructed in a manner that minimizes the potential for any freeze-thaw aging effects. The loss of material (spalling, scaling) and cracking due to freeze thaw are aging effects requiring management for PNPP Group 6 structures. PNPP is located in a severe weathering region per Figure 1 of American Society for Testing of Materials (ASTM) C33, Location of Weathering Regions, and group 6 structures are expected to experience the effects of freeze thaw. Accordingly, PNPP will monitor for loss of material (spalling, scaling) and cracking due to freeze thaw through the structures monitoring program. For further evaluation, see Section 3.5.2.2.2.3, Item 1 and 3.5.2.2.1.7.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 2 Page 4 of 6 LRA Section 3.5.2.2.2.3, Item 1 will be revised to state as follows:

The below grade inaccessible concrete areas of PNPP Group 6 structures were constructed in a manner that minimizes the potential for any freeze-thaw aging effects.

Therefore, even though loss of material (spalling, scaling) and cracking due to freeze-thaw is not expected, this aging effect for PNPP Groups 6 structures will still be managed under Structures Monitoring Program. The absence of concrete aging effects is confirmed by the Structures Monitoring and RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants programs.

ACI 318-71, Table 4.2-5 identifies the total air content percent by volume. Adequate quality controls were in place to reject concrete batches exceeding the air content percentage. For 3/4 aggregate, the allowable is between 4-8%. Concrete test reports from concrete pours used to construct various buildings indicated that all air content percentages are less than 8%. The foundation levels of all groups of structures are well below the frost line which would preclude this aging effect. Additionally, the 35 years of operating experience at PNPP has not resulted/indicated in any significant loss of material or cracking due to freeze thaw even in the accessible areas of concrete.

Accordingly, the fourth (last) paragraph of LRA Section 3.5.2.2.1.7 will be revised as follows:

From:

Therefore, loss of material (scaling, spalling) and cracking due to freeze thaw is not considered to be applicable to concrete in inaccessible areas. Concrete aging effects will continue to be monitored by the ASME Section XI, Subsection IWL for annulus concrete in containment and Structures Monitoring program for other buildings/locations.

To:

Concrete aging effects will continue to be monitored by the ASME Section XI, Subsection IWL for annulus concrete in containment and Structures Monitoring program for other buildings/ locations.

2. If LRA Item Number 3.5.1-49 is deemed applicable, identify an AMP to manage the aging effects, explain how the designated AMP will adequately manage freeze-thaw related aging effects in the inaccessible area of Group 6 concrete if they occur during the period of extended operation, and provide associate Table 2 item(s) in Table 3.5.2-3, Water Control Structures, accordingly. If LRA Item Number 3.5.1-49 is considered not applicable, provide a rationale for how inaccessible concrete areas that may be at risk for freeze-thaw cycles (e.g., portions of underground concrete walls located above or around the frost line) are excluded from aging management review.

LRA Item Number 3.5.1-49 is deemed applicable and as noted above in Item 1, Structures Monitoring program will manage this aging effect. The inaccessible areas will be managed as per the following existing enhancement as noted in the Structures Monitoring Program.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 2 Page 5 of 6 The program will be enhanced to require (a) evaluation of the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas and (b) examination of representative samples of the exposed portions of the below grade concrete, when excavated for any reason. If normally inaccessible areas become accessible due to planned activities, an inspection of these areas shall be conducted.

LRA Table 3.5.2-3 will have the following new rows added to address the aging effect loss of material (spalling, scaling) and cracking due to freeze-thaw in inaccessible concrete areas of Groups 6 Structures (Water-Control Structures):

LRA Table 3.5.2-3:

Component Type: Exterior walls below grade Intended Function: EN, HS, SNS, SSR,SRE Environment: Soil (Ext)

AERM: Cracking AMP: Structures Monitoring NUREG-1801 Item: III.A6.TP-110 Table 1 Item: 3.5.1-49 Note: G, w/Plant-specific Note as follows Plant-specific Note:

The environment is not listed in GALL for this component and material. AMR Table 1, 3.5.1, Item 3.5.1-49 was chosen to address this aging effect in the inaccessible areas of concrete for Group 6 Structures. A plant specific aging management program is not required because PNPP meets the conditions specified in the further evaluation section.

Structures Monitoring is the aging management program to manage this aging effect.

Component Type: Exterior walls below grade Intended Function: EN, HS, SNS, SSR,SRE Environment: Soil (Ext)

AERM: Loss of material AMP: Structures Monitoring NUREG-1801 Item: III.A6.TP-110 Table 1 Item: 3.5.1-49 Note: G, w/Plant-specific Note as noted above Component Type: Exterior walls above grade Intended Function: EN, HS, MB, SNS, SSR,SRE,FLB Environment: Air - outdoor (Ext)

AERM: Cracking AMP: Structures Monitoring NUREG-1801 Item: III.A6.TP-110 Table 1 Item: 3.5.1-49 Note: G, w/Plant-specific Note as noted above

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 2 Page 6 of 6 Component Type: Exterior walls above grade Intended Function: EN, HS, MB, SNS, SSR,SRE,FLB Environment: Air - outdoor (Ext)

AERM: Loss of material AMP: Structures Monitoring NUREG-1801 Item: III.A6.TP-110 Table 1 Item: 3.5.1-49 Note: G, w/Plant-specific Note as noted above Component Type: Floor slab Intended Function: EN, HS, SNS, SSR,SRE Environment: Raw water (Int)

AERM: Cracking AMP: Structures Monitoring NUREG-1801 Item: III.A6.TP-110 Table 1 Item: 3.5.1-49 Note: G, w/Plant-specific Note as noted above Component Type: Floor slab Intended Function: EN, HS, SNS, SSR,SRE Environment: Raw water (Int)

AERM: Loss of material AMP: Structures Monitoring NUREG-1801 Item: III.A6.TP-110 Table 1 Item: 3.5.1-49 Note: G, w/Plant-specific Note as noted above

3. Update the LRA accordingly based on the responses.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 3 Page 1 of 4 ESEB RAI-10327-R1 Regulatory Basis Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. As described in the SRP-LR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL-LR Report when evaluation of the matter in the GALL-LR Report applies to the plant.

Question 1 - TRP 45 AMP

Background

LRA Section B.2.43, as modified by LRA Supplement 3 (ML24206A150), provides an enhancement to the scope of program program element to include in-scope masonry walls for loss of material (spalling, scaling), change in material properties, and cracking due to freeze-thaw. The staff finds that this enhancement to the Structures Monitoring program contains both the scope of program and the parameters monitored or inspected program elements but lacks aging effect of cracking due to restraint shrinkage, creep, and aggressive environment.

AMR 3.5.1-70 in LRA Table 3.5.1 claims to be consistent with NUREG-1801 and contains aging effect of cracking due to restraint shrinkage, creep, and aggressive environment, which will be managed by the Structures Monitoring program.

AMR 3.5.1-71 in LRA Table 3.5.1 claims to be consistent with NUREG-1801 with aging effect of loss of material (spalling, scaling) and cracking due to freeze-thaw, which will be managed by the Structures Monitoring program.

Additionally, LRA Section B.2.43 states that the Masonry Walls program will be implemented under the Structures Monitoring program but lacks acceptance criteria for masonry walls. The acceptance criteria in G ALL-LR XI.S5 AMP states, [f]or each masonry wall, the extent of observed shrinkage and/or separation and cracking of masonry may not invalidate the evaluation basis or impact the walls intended function. However, further evaluation is conducted if the extent of cracking and loss of material is sufficient to impact the intended function of the wall or invalidate its evaluation basis.

Issue

1. The LRA lacks the enhancement to the parameters monitored or inspected and acceptance criteria program elements for in-scope masonry walls.

2. The LRA does not make clear whether cracking due to restraint shrinkage, creep, and aggressive environment needs to be included in the enhancement to the Structures Monitoring program.

3. The LRA lacks Table 2 items associated with AMR 3.5.1-71 for cracking.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 3 Page 2 of 4 Request

1. Clarify whether this enhancement to the Structures Monitoring program is for both the scope of program and the parameters monitored or inspected program elements.

2. Clarify whether this enhancement to the Structures Monitoring will include cracking due to restraint shrinkage, creep, and aggressive environment. If not, provide justification for why it is not applicable.

3. Provide Table 2 AMR items associated with AMR 3.5.1-71 for cracking.

4. Provide the enhancement to the acceptance criteria program element for masonry walls.

5. Revise the LRA accordingly based on the responses above.

PNPP Response

1. Clarify whether this enhancement to the Structures Monitoring program is for both the scope of program and the parameters monitored or inspected program elements.

Current LRA Section B.2.43 identifies the following enhancement under Scope Program Element 1:

The program implementing procedure will be enhanced to include the monitoring of in-scope non-safety related/non-seismic masonry walls for loss of material (spalling, scaling), change in material properties and cracking due to freeze-thaw.

LRA Sections A.1.43 and B.2.43 will be revised as follows:

Scope (Element 1)

The program implementing procedure will be enhanced to include the monitoring of in-scope non-safety related/non-seismic masonry walls for aging management.

Parameters Monitored/Inspected (Element 3)

The program implementing procedure will be enhanced to include the monitoring of in-scope masonry walls for loss of material (spalling, scaling), change in material properties and cracking due to freeze-thaw.

2. Clarify whether this enhancement to the Structures Monitoring will include cracking due to restraint shrinkage, creep, and aggressive environment. If not, provide justification for why it is not applicable.

Enhancements to Structures Monitoring Program will not include cracking due to restraint shrinkage, creep, and aggressive environment due to the following reasons:

The effects of creep are small and not consequential to the intended functions of masonry walls. Creep does occur early in the concrete life; however, no cracks have been observed.

Shrinkage in masonry walls is not a long-term aging mechanism and is not expected to continue after 40 years and into the LR period.

The PNPP outside air environment is non-aggressive per LRA 3.6.2.2.2. PNPP is located in an area with moderate rainfall and where the outdoor environment is not subject to industry

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 3 Page 3 of 4 air pollution or salt spray. The in-scope masonry walls are not located below grade.

Therefore, masonry walls are not subject to an aggressive environment.

LRA Table 3.5.1, Item 3.5.1-70 will be revised to state as follows:

Not applicable - The masonry walls that are within the scope of license renewal at PNPP are limited to isolated non-safety related, non-seismic Category I structures not meeting the criteria of I.E Bulletin 80-11. The in-scope masonry walls do not include integrated restraints and are not subject to restraint shrinkage. The effects of creep are small and not considered consequential to the intended functions of masonry walls. PNPP is located in an area with moderate rainfall and where the outdoor environment is not subject to industry air pollution or salt spray.

Additionally, LRA Table 3.5.2-2 will be revised as follows:

Rows 129, 132, 134 and 137 of LRA Table 3.5.2-2 (Component Type Masonry walls, Material Concrete Block) will be deleted since each of those rows corresponds to NUREG 1801 Item III.A3.TP-12 involving the aging effect/mechanism of cracking due to restraint shrinkage, creep, and aggressive environment.

3. Provide Table 2 AMR items associated with AMR 3.5.1-71 for cracking.

PNPP concurs with NRC recommendation. LRA Table 3.5.2-2 will have the following row added for masonry walls to address cracking based on Table 1 Item 3.5.1-71:

Component Type: Masonry Walls Intended Function: SNS, SRE Material: Concrete block Environment: Air - outdoor (Ext)

AERM: Cracking AMP: Structures Monitoring NUREG-1801 Item: III.A5.TP-34 Table 1 Item: 3.5.1-71 Note: E

4. Provide the enhancement to the acceptance criteria program element for masonry walls.

PNPP concurs with NRC recommendation. LRA Sections A.1.43 and B.2.43 will be updated to add the following enhancement.

Acceptance Criteria (Element 6)

Implementing procedures will be updated to reflect when cracking or separation are observed in in-scope masonry walls, a condition report shall be initiated to document an evaluation of the effect of the condition for acceptability on the intended function of the masonry wall.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 3 Page 4 of 4

5. Revise the LRA accordingly based on the responses above.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 4 Page 1 of 5 ESEB RAI-10327-R1 Question 2 - TRP 46 AMP

Background

LRA Section B.2.43, as modified by LRA Supplement 3 (ML24206A150), provides an enhancement to the parameters monitored or inspected program element to monitor the porous sub-foundation for loss of material and change in material properties.

Table 2 AMR items in LRA Table 3.5.2-1 (items 23 and 26), as modified by LRA Supplement 3 (ML24206A150), adds additional aging effect of increase in porosity and permeability, and loss of strength for porous concrete foundation. Additionally, based on the staff's assessment, cracking due to erosion of porous concrete sub-foundation may be an applicable aging effect.

Table 2 AMR item in LRA Table 3.5.2-1 (item 25), as modified by LRA Supplement 3 (ML24206A150), lists porous concrete sub-foundation for change in material properties, which is managed by the Structures Monitoring program, citing Note G with GALL-LR item II.B3.1.C-07 and AMR 3.5.1-2.

Issue

1. The LRA has inconsistent aging effects between the enhancement to the Structures Monitoring program and Table 2 AMR items.

2. The LRA lacks the detection of aging effects and the acceptance criteria program elements for the porous concrete sub-foundation and porous concrete pipe associated with the plant underdrain system.

3. Table 2 AMR item in LRA Table 3.5.2-1 (item 25), citing Note G, should not have any GALL-LR item II.B3.1.C-07 and AMR 3.5.1-2 associated with this line item.

Request

1. Evaluate and clarify aging effects of the porous concrete sub-foundation and porous concrete pipe associated with the plant underdrain system.

2. Clarify whether GALL-LR item B3.1.C-07 and AMR 3.5.1-2 in LRA Table 3.5.2-1 (item 25) can be deleted.

3. Revise the enhancement to the Structures Monitoring program and Table 2 AMR items for the porous concrete sub-foundation accordingly to ensure each Table 2 AMR item has the corresponding aging effect for the porous sub-foundation.

4. Provide the enhancement to the Structures Monitoring program and the corresponding Table 2 AMR items for the porous concrete pipe.

5. Evaluate whether the enhancements to the detection of aging effects and acceptance criteria program elements are needed for the porous concrete sub-foundation and porous concrete pipe associated with the plant underdrain system. If not, provide the justification for why they are not needed.

6. Revise the LRA accordingly based on the responses above.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 4 Page 2 of 5 PNPP Response

1. Evaluate and clarify aging effects of the porous concrete sub-foundation and porous concrete pipe associated with the plant underdrain system.

PNPP concurs with NRC Staffs comment that there are inconsistencies between the LRA Section B.2.43 and Table 2 items associated with porous concrete sub-foundation. PNPP also concurs with NRCs recommendation to add aging effect cracking due to erosion of porous concrete subfoundation to LRA Tables 3.5.2-1 and 3.5.2-2. Also see response to Request 5 below on porous concrete pipe.

The following changes will be made to the LRA to optimize consistency between the LRA Section B.2.43 and Table 2 items associated with the porous concrete sub-foundation.

New rows will be added to LRA Tables 3.5.2-1, 3.5.2-2 and 3.5.2-3 associated with the Component Type porous concrete sub-foundation as follows:

LRA Table 3.5.2-1:

Component Type: Containment foundation Intended Function: SNS, SRE, SSR Material: Porous concrete Environment: Raw water (Int)

AERM: Reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub-foundation AMP: Structures Monitoring Program NUREG-1801 Item: II.B3.1.C-07 Table 1 Item: 3.5.1-2 Note: A Component Type: Containment foundation Intended Function: SNS, SRE, SSR Material: Porous concrete Environment: Soil (Ext)

AERM: Reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub-foundation AMP: Structures Monitoring Program NUREG-1801 Item: II.B3.1.C-07 Table 1 Item: 3.5.1-2 Note: A LRA Table 3.5.2-2:

Component Type: Foundations Intended Function: EN, SRE, SSR Material: Porous concrete Environment: Raw water (Int)

AERM: Reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub-foundation AMP: Structures Monitoring Program NUREG-1801 Item: III.A3.TP-31

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 4 Page 3 of 5 Table 1 Item: 3.5.1-46 Note: A Component Type: Foundations Intended Function: EN, SRE, SSR Material: Porous concrete Environment: Soil (Ext)

AERM: Reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub-foundation AMP: Structures Monitoring Program NUREG-1801 Item: III.A3.TP-31 Table 1 Item: 3.5.1-46 Note: A LRA Table 3.5.2-3:

Component Type: Foundation Intended Function: EN, HS, SNS, SSR Material: Porous concrete Environment: Raw water (Int)

AERM: Reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub-foundation AMP: Structures Monitoring Program NUREG-1801 Item: III.A6.TP-31 Table 1 Item: 3.5.1-46 Note: A Component Type: Foundation Intended Function: EN, HS, SNS, SSR Material: Porous concrete Environment: Soil (Ext)

AERM: Reduction of foundation strength and cracking due to differential settlement and erosion of porous concrete sub-foundation AMP: Structures Monitoring Program NUREG-1801 Item: III.A6.TP-31 Table 1 Item: 3.5.1-46 Note: A LRA Sections A.1.43 and B.2.43, enhancements for Parameters Monitored/Inspected (Element

3) will be updated as follows:

From:

(Parameters Monitored/Inspected, Element 3)

The program will be enhanced to monitor the porous concrete sub-foundation for:

Loss of material (erosion of porous concrete sub-foundation)

Change in material properties (leaching of calcium hydroxide)

Monitoring the building settlement

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 4 Page 4 of 5 To:

(Parameters Monitored/Inspected, Element 3)

The implementing procedures will be enhanced to monitor the porous concrete sub-foundation for:

Loss of material (erosion of porous concrete sub-foundation)

Change in material properties (leaching of calcium hydroxide)

Increase in porosity and permeability, loss of strength Reduction of foundation strength and cracking due to differential settlement and erosion of the porous concrete sub-foundation

2. Clarify whether GALL-LR item B3.1.C-07 and AMR 3.5.1-2 in LRA Table 3.5.2-1 (item

25) can be deleted.

PNPP concurs with NRC Staffs assertion there is an issue in this line item. Standard Note G will be replaced with Standard Note A. The aging effect Change in material properties is an applicable aging effect for porous concrete based upon the PNPPs bases document. The row associated with Component Types "Containment foundation" in Table 3.5.2-1, "Foundations" with aging effect Change in material properties for porous concrete will be revised as follows:

LRA Table 3.5.2-1:

Component Type: Containment foundation Intended Function: SNS, SRE, SSR Material: Porous concrete Environment: Soil (Ext)

AERM: Change in material properties AMP: Structures Monitoring Program NUREG-1801 Item: II.B3.1.C-07 Table 1 Item: 3.5.1-2 Note: A

3. Revise the enhancement to the Structures Monitoring program and Table 2 AMR items for the porous concrete sub-foundation accordingly to ensure each Table 2 AMR item has the corresponding aging effect for the porous sub-foundation.

The responses to requests 1 and 2 above detail the proposed changes to the PNPP LRA.

4. Provide the enhancement to the Structures Monitoring program and the corresponding Table 2 AMR items for the porous concrete pipe.

Section 2.3.3.44 of LRA which is based on PNPP UFSAR Section 2.4.13.5 Pressure Relief Underdrain System Description, states that, the underdrain system consists of a porous concrete blanket, nominally one foot thick, which underlies all of the structures of the nuclear island. Between some of the buildings and around the perimeter of the nuclear island, the blanket is increased in thickness to incorporate a one foot diameter, porous concrete pipe.

Based on the above description, for the purpose of aging management, the porous concrete pipe is considered integral to porous concrete sub-foundation, therefore no additional Table 2 items have been provided for porous concrete pipe.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 4 Page 5 of 5

5. Evaluate whether the enhancements to the detection of aging effects and acceptance criteria program elements are needed for the porous concrete sub-foundation and porous concrete pipe associated with the plant underdrain system. If not, provide the justification for why they are not needed.

PNPP concludes that any additional enhancements to the detection of aging effects and acceptance criteria program elements are not needed for the porous concrete sub-foundation and the porous concrete pipe associated with the plant underdrain system due to the following reasons:

Plant implementing procedures are currently in place that monitor the porous concrete subfoundation and its performance to assure that it is performing its intended function. They include periodic monitoring of groundwater elevation, cleaning, inspection and maintenance of porous concrete pipe, monitoring of any erosion from the porous concrete sub-foundation, e.g.,

presence of any aggregates in the manholes, and monitoring of any building settlement and appearance of any cracks in the building foundations. The frequencies of this type of monitoring as documented in the plant implementing procedures is less than the required five-year monitoring of building inspections. The current monitoring procedures will be in place during the period of extended operation. Therefore, no additional enhancements to the detection of aging effects and acceptance criteria program are needed.

6. Revise the LRA accordingly based on the responses above.

LRA changes associated with this RAI response will be provided in a future LRA supplement.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 5 Page 1 of 3 ESEB RAI-10327-R1 Question 3 - TRP 46 AMP

Background

LRA Section B.2.43, as modified by LRA Supplement 3 (ML24206A150), provides an enhancement to the parameters monitored or inspected program element to monitor the plant drain piping for unacceptable flow blockage, change in internal geometry, or other internal degradation irrespective of piping material.

Table 2 AMR items in LRA Table 3.5.2-4 list steel and concrete storm drain (items 322 to 324, items 327 to 328) for loss of material and flow blockage, polymer storm drain (item 325) for flow blockage, which will be managed by the Structures Monitoring.

Issues

1. The aging effects of the plant storm drain in the enhancement to the Structures Monitoring program are inconsistent with ones for Table 2 AMR items.

2. The LRA lacks Table 2 AMR item for loss of material of the polymer storm drain.

Request

1. Clarify the aging effects of the plant drain piping and ensure all aging effects are identified.

2. Revise the enhancement to the Structures Monitoring program and Table 2 AMR items accordingly to ensure each Table 2 AMR item has the corresponding aging effect for the plant drain piping based on the responses above.

PNPP Response

1. Clarify the aging effects of the plant drain piping and ensure all aging effects are identified.

The storm drain system is described in PNPPs LRA Section 2.3.3.58. That LRA section indicates the P67 storm drain and sewer system components are scoped under structural as structural bulk commodities. Structural bulk commodities are described in PNPPs LRA Section 2.4.4. The bulk commodities associated with storm drains identified in LRA Table 2.4.4-1 include:

Catch Basins and Catch basin covers; and Storm Drains (Steel, Polymer and Concrete)

A review of LRA AMR Table 3.5.2-4 for Bulk Commodities for the above storm drain components includes the following components (materials):

Catch Basin Steel Covers (steel)

Catch Basin Steel Covers1 (steel)

Catch Basins and Catch Basin Covers (concrete)

Storm drain (steel)

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 5 Page 2 of 3 Storm Drain 1 (polymer)

Storm Drain 3 (concrete)

Thus, there is consistent alignment between LRA Sections 2.3.3.58, 2.4.4 and Table 3.5.2-4 for identified storm drain components and materials.

LRA Table 3.5.2-4 identifies the following aging effects that are monitored:

Flow blockage (for steel, polymer and concrete storm drains)

Loss of material (for steel and concrete storm drains)

Loss of material (for steel catch basin covers)

Change in material properties for concrete catch basins and catch basin covers Cracking for concrete catch basins and catch basin covers Loss of material (Corrosion of embedded steel reinforcing)

Based on the above, it is evident that all of the aging effects applicable to storm drain piping (concrete) were not cited in LRA Table 3.5.2-4. In order to resolve this issue, it is proposed that the following changes are applied to LRA Table 3.5.2-4 for the component Storm Drain3 (concrete):

Component Type: Storm Drain3 Intended Function: FLB SNS Material: Concrete Environment: Air Outdoor (int)

AERM: Cracking/Loss of material Aging Management Program: Structures Monitoring NUREG 1801 Item: III.A3.TP-23 Table 1 Item: 3.5.1-64 Notes: A Component Type: Storm Drain3 Intended Function: FLB SNS Material: Concrete Environment: Raw water (int)

AERM: Loss of material (Corrosion of embedded steel reinforcing)

Aging Management Program: Structures Monitoring NUREG 1801 Item: III.A3.TP-212 Table 1 Item: 3.5.1-65 Notes A In addition to monitoring for the cited aging effects for storm drain piping as identified in Table 3.5.2-4, the Structures monitoring aging management program includes the following existing enhancement in LRA sections A.1.43 and B.2.43:

Parameters Monitored/Inspected (Element 3)

The program will be enhanced to require that plant storm drain piping is monitored for:

Unacceptable flow blockage, change in internal geometry, or other internal degradation irrespective of piping material.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 5 Page 3 of 3 During opportunistic excavations of piping:

external corrosion in steel (corrugated metal) pipe reflecting loss of material, external spalled concrete, external degradation of polymer piping such as blistering or delamination Regarding the aging effects for the storm drain1 (polymer) component in LRA Table 3.5.2-4, the polymer applied in the storm drain piping is limited to rigid polymers, including PVC (polyvinyl chloride) and HDPE.

The storm drain piping is basically gravity pipe subject to very little internal pressure. Flow in storm drain piping is mostly non-existent and when there is flow it is low velocity and not abrasive. Generally, buried storm drain piping of all materials have a service life ranging from 50 to 100 years and have been performing their intended function with little or no challenges. Since storm drain piping does not perform a pressure boundary function, application of traditional wall thickness monitoring is not considered applicable. The presence of cracks, holes or perforations does not affect the ability of the storm drain piping to perform its intended function, i.e., flow of water.

Additionally, loss of material in plastic piping is not discussed as an applicable aging effect in EPRI Mech Tools, App E, Section 3.6.2 (EPRI Report 1010639). PNPP uses EPRIs Mechanical Tools, Appendix A, Section 3.6.2, for evaluating aging effects which identifies there are no loss of material as an aging effect. Based upon these facts, PNPP noted that the surfaces are:

not exposed to temperatures exceeding 150 degrees F (considered the threshold temperature for PVC),

not exposed to aggressive chemicals, and not exposed to Ultraviolet light, ozone, or ionizing radiation.

Therefore, no aging effects beyond flow blockage for polymer storm drain piping are considered to be warranted. Again, the enhancement cited in LRA section B.2.43 for monitoring storm drain piping is considered to manage any potential unforeseen age-related degradation.

2. Revise the enhancement to the Structures Monitoring program and Table 2 AMR items accordingly to ensure each Table 2 AMR item has the corresponding aging effect for the plant drain piping based on the responses above.

Based on the response to Request 1 above, changes to AMR Table 3.5.2-4 are identified to manage the aging effects of concrete storm drain piping. No changes to the enhancements to the Structures Monitoring aging management program regarding storm drain components were determined to be necessary.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 6 Page 1 of 2 ESEB RAI-10327-R1 Question 4 - TRP 46 AMP

Background

GALL-LR XI.S6 AMP states that the scope of this program includes periodic sampling and testing of ground water. LRA Section B.2.43, as modified by LRA Supplement 3 (ML24206A150), provides an enhancement to the parameters monitored or inspected program element to monitor ground water chemistry.

Issue The LRA lacks an enhancement to the scope of program program element for the periodic sampling.

Request Clarify whether an enhancement to the scope of program program element is needed for the periodic sampling and testing of ground water. If not, provide the justification for why it is not needed.

PNPP Response Clarify whether an enhancement to the scope of program program element is needed for the periodic sampling and testing of ground water. If not, provide the justification for why it is not needed.

The PNPP LRA Section B.2.43 program description includes the following 2 statements:

The scope of the structures monitoring program also includes managing aging effects associated with the plant underdrain system, storm drains that are considered as bulk commodities and porous concrete sub-foundations.

The program also includes provisions for periodic testing and assessment of ground water chemistry and inspection of accessible below grade concrete structures.

Additionally, as part of the changes submitted under PNPP LRA Supplement 3 (Vistra Letter L-24-108, Attachment 36) the following enhancement was also provided:

Groundwater chemistry parameters will be monitored on a frequency of at least once every five years. (Element 4)

Corresponding acceptance criteria was added to the program via PNPP LRA Supplement 3 outlining when groundwater water would no longer be considered non-aggressive. (Element 6)

Based on this description of the scope of the structures monitoring program, an additional enhancement addressing periodic sampling and testing of ground water does not appear to be warranted.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 6 Page 2 of 2 There is no change to the LRA for this RAI response.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 7 Page 1 of 5 ESEB RAI-10327-R1 Question 5 - TRP 46 AMP

Background

LRA Section B.2.43, as modified by LRA Supplement 3 (ML24206A150), provides a new enhancement to the acceptance criteria program element to monitor unimpregnated and impregnated (with elastomer) fiberglass fabric for loss of material, cracking/delamination, change in material properties and visual deterioration. In addition, LRA Section 2.4.4 Structural bulk commodities does not include unimpregnated and impregnated (with elastomer) fiberglass fabric.

There is only one Table 2 AMR item in LRA Table 3.5.2-4 (item 314), which addresses the change in material properties and cracking for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer, which are managed by the Structures Monitoring program, citing Note J.

Additionally, it is noted that the aging effects of the other related Table 2 AMR items in LRA Table 3.5.2-4 (items 318 and 319) that have the same combination of material, aging effect and environment are managed by the Fire Protection program citing Note H instead of the Structures Monitoring program.

Issues

1. It appears that this enhancement to the Structures Monitoring program does not belong to the acceptance criteria program element. In addition, LRA Section 2.4.4 Structural bulk commodities does not include unimpregnated and impregnated (with elastomer) fiberglass fabric.

2. The LRA does not make clear why different Notes J and H ( for the Structures Monitoring and the Fire Protection programs, respectively) are used for the same combination of material, aging effect and environment.

3. The LRA lacks Table 2 AMR item for loss of material of unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer managed by the Structures Monitoring program.

Request

1. Clarify whether this enhancement to the Structures Monitoring program is applicable to both the scope of program and the parameters monitored or inspected program elements.

2. Explain why both Note J and Note H ( for the Structures Monitoring and the Fire Protection programs, respectively) are used for the same combination of material, aging effect, and environment, and correct them if necessary.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 7 Page 2 of 5

3. Evaluate the aging effects in the enhancement to the Structures Monitoring and ensure all applicable aging effects are identified and are consistent with aging effects in the Table 2 AMR items.

4. Revise the enhancement to the Structures Monitoring program and Table 2 AMR items accordingly to ensure each Table 2 AMR item has the corresponding aging effect for the plant drain piping based on the responses above.

PNPP Response

1. Clarify whether this enhancement to the Structures Monitoring program is applicable to both the scope of program and the parameters monitored or inspected program elements.

The enhancement to Structures Monitoring Program, Acceptance Criteria (Element 6) (as modified by Supplement 3 to the PNPP LRA (Vistra Letter L-24-108) included the following:

The program will be enhanced to monitor unimpregnated and impregnated (with elastomer) fiberglass fabric for loss of material, cracking/delamination, change in material properties and visible deterioration.

PNPP concurs that the above enhancement does not fall under the category Acceptance Criteria (Program Element 6). It actually belongs to both Program Elements 1 and 3.

Therefore, the LRA Section B.2.43 will be revised to delete the above enhancement from Acceptance Criteria (Program Element 6) and added to both Program Elements 1 and 3.

Specifically, the following enhancements will be added as follows:

Scoping (Element 1)

The program implementing procedures will be enhanced to monitor unimpregnated and impregnated (with elastomer) fiberglass fabric for aging effects.

Parameters Monitored/Inspected (Element 3).

The program implementing procedures will be enhanced to monitor unimpregnated and impregnated (with elastomer) fiberglass fabric for loss of material, separation, cracking/delamination, and change in material properties and visible deterioration.

Note: An additional aging effect Separation was added in the above enhancement for this material.

2. Explain why both Note J and Note H ( for the Structures Monitoring and the Fire Protection programs, respectively) are used for the same combination of material, aging effect, and environment, and correct them if necessary.

The updated Table 3.5.2-4 for Bulk Commodities as identified in LRA Supplement 3 (PNPP Letter L-24-108), dated July 24, 2024, Attachment 18 identifies the following rows where the material Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer has been used. They are identified in Rows 262, 263, 314, 318 and 319. It should be noted that with the exception of Row 314 which is for shielding, Fire Protection

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 7 Page 3 of 5 Program with Note H is the aging management program used for managing the effects for these component types that have a fire barrier function. On the contrary, Row 314 is a component type providing radiation shielding with no fire barrier function, consequently Structures Monitoring Program with Note J has been used. Therefore, use of Notes H and J as noted above is appropriate.

3. Evaluate the aging effects in the enhancement to the Structures Monitoring and ensure all applicable aging effects are identified and are consistent with aging effects in the Table 2 AMR items.

Considering all the aging effects associated with this material, and maintaining consistency between LRA Section B.2.43 and Table 3.5.2-1 and Table 3.5.2-4, the following actions will be taken:

LRA Table 3.5.2-4, will have two additional aging effects Cracking/Delamination and Separation added as noted below for the Component Types noted:

Table 3.5.2-4 Component Type: Penetration sealant (fire)

Intended Function: FB, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking/Delamination AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-4 Component Type: Penetration sealant (fire)

Intended Function: FB, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-4 Component Type: SRV Tailpipe Penetration Boot Seals Intended Function: EN, FB, SPB Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking/Delamination AMP: Fire Protection

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 7 Page 4 of 5 NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-4 Component Type: SRV Tailpipe Penetration Boot Seals Intended Function: EN, FB, SPB Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 LRA Table 3.5.2-1, will have two additional aging effects Cracking/Delamination and Separation added as noted below for the Component Type noted:

Table 3.5.2-1 Component Type: Drywell mechanical penetration (fiberglass fabric)

Intended Function: FB, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking/Delamination AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-1 Component Type: Drywell mechanical penetration (fiberglass fabric)

Intended Function: FB, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Note: The above response also addresses answers to Question 4 associated with Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer in NCSG RAI-10337-R1.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 7 Page 5 of 5

4. Revise the enhancement to the Structures Monitoring program and Table 2 AMR items accordingly to ensure each Table 2 AMR item has the corresponding aging effect for the plant drain piping based on the responses above Enhancements to Structures Monitoring Program are addressed in Item 1 above. AMR Table 3.5.2-4 will be revised as noted in Item 3 above.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 8 Page 1 of 2 ESEB RAI-10328-R1 Regulatory Basis Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. As described in the SRP-LR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL-LR Report when evaluation of the matter in the GALL-LR Report applies to the plant.

Question 1 -TRP 46 AMR

Background

Table 2 AMR item in LRA Table 3.5.2-4 (item 77), as modified by Supplement (ML24206A150),

addresses loss of strength for polymer conduit caps exposed to air-indoor environment, which will be managed by the Structures Monitoring program. This Table 2 AMR item cites plant-specific notes 518, which states in part, "[t]he function of the conduit cap is an external flood barrieras long as the conduit cap is in place and shows minimal signs of deterioration, it is capable of performing its intended function of providing an adequate seal for flooding." Based on the staff assessment, it appears that loss of material, loss of seal, and cracking are applicable aging effects of the polymer conduit caps.

Issue The LRA does not make clear whether all applicable aging effects such as loss of material, loss of sealing, and cracking are identified.

Request Clarify whether loss of material, loss of sealing, and cracking are applicable aging effects for the polymer conduit caps. If not, provide the justification for why they are not applicable. Otherwise, provide Table 2 AMR items associated with their aging effects along with aging mechanisms.

PNPP Response Clarify whether loss of material, loss of sealing, and cracking are applicable aging effects for the polymer conduit caps. If not, provide the justification for why they are not applicable. Otherwise, provide Table 2 AMR items associated with their aging effects along with aging mechanisms.

The aging effects cited in the request, e.g., loss of material, loss of sealing, and cracking are considered not applicable aging effects for the polymer conduit caps due to the following reasons:

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 8 Page 2 of 2 Conduit caps at PNPP are comprised of a rigid polymer (PVC). PNPP uses EPRIs Mechanical Tools, Appendix A, Section 3.6.2, for evaluating aging effects and concluded there are no aging effects, including loss of material, loss of sealing, or cracking. The conduit caps are located inside buildings. Based upon these facts, PNPP noted that the polymer conduit caps are not exposed to:

temperatures exceeding 150 degrees F (considered the threshold temperature for PVC),

aggressive chemicals, and ultraviolet light, ozone, or ionizing radiation.

LRA Table 3.5.2-4 conservatively assigns an aging effect of loss of strength in order for these conduit caps to be subject to monitoring under the structures monitoring program.

The associated plant specific note for Table 3.5.2-4 will be modified as shown below:

The function of the conduit cap is to protect conduits from flooding. Evaluation has concluded no aging effects apply including cracking, loss of material, or loss of sealing. Structures monitoring will monitor for evidence of deterioration to ensure the intended function of providing an adequate seal for flooding. These PVC conduit caps are not expected to experience aging effects because they are not exposed to elevated temperatures above 150 degrees F, ozone, nor ultraviolet or ionizing radiation. These components are located within structures.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 9 Page 1 of 2 ESEB RAI-10328-R1 Question 2 TRP 46 AMR

Background

Table 2 AMR item in LRA Table 3.5.2-4 (item 284), as modified by LRA Supplement 3 (ML24206A150), appears to list an incorrect Table 1 AMR item 3.5.1-33 (GALL-LR item II.B4.CP-41) for the elastomer roof membranes. AMR 3.5.1-33 (GALL-LR item II.B4.CP-41) is for component in containment structures, which is managed by the 10 CFR Appendix J program.

Table 2 AMR items in LRA Table 3.5.2-4 (items 353, 354 355), as modified by LRA Supplement 3 (ML24206A150), lists cracking for the elastomer waterproofing & membranes, which are managed by the Structures Monitoring program.

GALL-LR item III.A6.TP-7 addresses loss of sealing due to deterioration of seals, gaskets, and moisture barriers (caulking, flashing, and other sealants), which is managed by the Structures Monitoring program. Loss of sealing is an applicable aging effect for elastomer roof membranes and waterproofing & membranes.

Issue The LRA does not address loss of sealing for elastomer roof membranes and waterproofing &

membranes.

Request

1. Clarify whether loss of sealing is an applicable aging effect, if not, provide the justification for why it is not applicable. Otherwise, provide Table 2 AMR items for the elastomer roof membranes and waterproofing & membranes.

2. Clarify whether 5.1-33/II.B4.CP-41 is the appropriate Table 2 AMR item in LRA Table 3.5.2-4 (item 284) and revise Note E accordingly.

PNPP Response

1. Clarify whether loss of sealing is an applicable aging effect, if not, provide the justification for why it is not applicable. Otherwise, provide Table 2 AMR items for the elastomer roof membranes and waterproofing & membranes.

Yes, loss of sealing is an applicable aging effect. PNPP acknowledges that this nomenclature is not in LRA Table 3.5.2-4 for elastomer roof membranes and waterproofing & membranes.

However, it is included in the meaning of change in material properties and could manifest itself via the aging effect cracking depending on its severity.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 9 Page 2 of 2 Per EPRI 3002013084, Structural Tools, Table 7-5, the aging effect of Loss of sealing for elastomers in Air - outdoor, raw water, and concrete share the same aging mechanisms:

Thermal exposure, UV radiation and ozone. Concrete and Raw water include ionizing radiation.

None of these mechanisms were evaluated as applying to these component types and material combinations since the thresholds for these aging mechanisms were not exceeded.

Nevertheless, loss of sealing results from leakage in elastomer roof membrane assemblies/composite materials from two other aging effects change of material properties and cracking. Therefore, although LRA Table 3.5.2-4 does not specifically include Loss of sealing, its effects are manifested through these other aging effects. This condition is described in the LRA. In the fourth paragraph in LRA Section B.2.43 (page B-120) as modified by LRA Supplement 3 (Vistra Letter L-24-108 and ML24206A150), the meaning of the aging effect Change in material properties, as it applies to elastomers, includes hardening, shrinkage, and loss of sealing. Additionally, this situation is specifically included in the Discussion text in Table 3.5.1, line item 3.5.1-33.

Since this approach is handled consistently via the bases document and included in the LRA, it is concluded that changes to the Table 2 AMR for roof membranes and waterproofing membranes are not warranted.

2. Clarify whether 5.1-33/II.B4.CP-41 is the appropriate Table 2 AMR item in LRA Table 3.5.2-4 (item 284) and revise Note E accordingly.

PNPPs license renewal evaluation process focuses on alignment with materials, environments, and aging effects (MEA) that are in NUREG-1801 and yet remains consistent with the information in EPRI tools (mechanical and structural). Consequently, PNPP personnel will use line items from other chapters when there is no matching MEA in the NUREG-1801 chapter containing the building and component type being addressed. In this case item CP-41 in NUREG-1801 only appears in Chapter II, Tables A3 and B4. The MEA for Item CP-41 is Elastomers, rubber and other similar materials exposed to Air - indoor, uncontrolled or Air -

outdoor environments resulting in Loss of sealing due to wear, damage, erosion, tear, surface cracks, or other defects. In this case, change in material and cracking were associated with other defects. Note E was used since the Structures Monitoring program will manage these aging effects/ mechanisms and not the PNPP equivalent aging management program to Chapter XI.S4, 10 CFR Part 50, Appendix J."

No changes to the LRA are necessary.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 10 Page 1 of 2 0 ESEB RAI-10328-R1 Question 3 TRP 46 AMR

Background

AMR 3.5.1-63 in LRA Table 3.5.1 claims to be consistent with NUREG-1801. However, Table 2 AMR items associated with Table 1 AMR item 3.5.1-63 in LRA Tables 3.5.2-1 (items 50 and 147) and 3.5.2-2 (items 25 37, 56, and 169), as modified by LRA Supplement 3 (ML24206A150), list raw water and soil environments, which will be managed by the Structures Monitoring program.

However, their corresponding GALL-LR items III.A1.TP-24 and III.A3.TP-24 list water-flowing environments.

Issue Table 2 AMR items in LRA Tables 3.5.2-1 (items 50 and 147) and 3.5.2-2 (items 25 37, 56, and 169), as modified by LRA Supplement 3 (ML24206A150), are inconsistent with GALL-LR items III.A1.TP-24 and III.A3.TP-24 in the environment.

Request Address the discrepancy in the environment and revise Table 2 AMR items with the correct Note accordingly.

PNPP Response Address the discrepancy in the environment and revise Table 2 AMR items with the correct Note accordingly.

PNPP does not consider there to be a discrepancy in the identified environments.

As shown in the following excerpt for Table 3.0-1 of the PNPP LRA, the definitions for the raw water and soil environments include flowing water:

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 10 Page 2 of 2 PNPP LRA Table 3.0-1 PNPP Service Environments for Mechanical and Structural Aging Management Reviews AMR Environment NUREG 1801 Environment(s) used for AMR comparison [1,2]

Definition Raw water Ground water; Raw water; Water, flowing or standing Raw, untreated, river, lake, ground or potable water.

Raw water does not exceed the threshold temperature for SCC of stainless steels (140°F). If raw water is considered aggressive to structural components, additional aging management methods may be required.

Soil Soil; Water - flowing under foundation; Water - flowing or standing External environment for components exposed to soil (including the air/soil interface) or buried in the soil, including groundwater in the soil. This name is also used to describe the environment for exterior surface of outdoor tank bottoms that are mounted on a concrete pad. Buried components are not expected to exceed the threshold temperature for SCC (140°F).

These definitions are consistent with NUREG-1801, Chapter IX, Table IX.D, which includes the following definition for Water-flowing:

Water-flowing - Water that is refreshed; thus, it has a greater impact on leaching and can include rainwater, raw water, ground water, or water flowing under a foundation.

The evaluation of the aging effects assigned to the components in the raw water or soil environment was consistent with the GALL flowing water environment unless individually noted.

No changes to the LRA were considered to be necessary due to this request.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 11 Page 1 of 4 1 NCSG RAI-10332-R1 Regulatory Basis Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S. Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a),

the staff requires additional information in regard to the matters described in the requests for information.

Question 1 Coolant Heat Exchanger Tube Bundle Replacement Frequency

Background

Supplement 2 of the License Renewal Application (LRA) dated June 27, 2024 (ML24180A010),

revised LRA Section A.1.21, LRA Table A.3, and LRA Section B.2.21 to include an enhancement to the Fire Water System program to periodically replace the coolant heat exchanger tube bundle on the diesel driven fire pump engine during the period of extended operation. The enhancement included a periodic replacement frequency of every 14 years.

Supplement 2 also revised LRA Table 3.3.2-24 to manage loss of material of the diesel fire pump heat exchanger shell, and loss of material and reduction of heat transfer of the diesel fire pump heat exchanger tubes exposed to closed-cycle cooling water by the Fire Water System program in lieu of the Closed Treated Water Systems program.

During the audit of the Fire Water System program, several procedures were discussed.

Specifically, PAP-1910, Fire Protection Program, that requires the diesel driven fire pump be operated every 31 days, including inspection for leaks and condition of all piping and component types containing coolant, and leakage checks every 3 months; and PMI-0072, Diesel Fire Service Pump Preventive Maintenance, that requires the coolant to be drained and replaced annually, and components to be inspected for mineral buildup, scale, rust, or oil, and cleaned as needed. It was also noted that coolant temperature is monitored to confirm heat transfer performance while the engine is running.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 11 Page 2 of 4 Issue Supplement 2 of the LRA did not provide a technical basis for the periodic replacement frequency of every 14 years. It is unclear whether the diesel driven fire pump engine monitoring and maintenance activities in PAP-1910 and PMI-0072 will continue during the period of extended operation, and whether the activities are equivalent to the activities that would be performed under the Closed Treated Water Systems program. For example, depending on the specific water chemistry treatment program, the Closed Treated Water Systems program includes quarterly chemistry testing.

Request

1. Please provide the technical basis supporting the periodic replacement frequency of every 14 years for the coolant heat exchanger tube bundle on the diesel driven fire pump engine.

2. Please discuss whether the diesel driven fire pump engine monitoring and maintenance activities in PAP-1910 and PMI-0072 will continue during the period of extended operation, including before and after replacement of the coolant heat exchanger tube bundle.

3. Please provide sufficient information to demonstrate that the diesel driven fire pump engine monitoring and maintenance activities that will be performed by the Fire Water System program are or will be (i.e., require program enhancement(s)) equivalent to the monitoring and maintenance activities that would be performed by the Closed Treated Water Systems program.

PNPP Response

1. Please provide the technical basis supporting the periodic replacement frequency of every 14 years for the coolant heat exchanger tube bundle on the diesel driven fire pump engine.

PNPP is not aware of any tube bundle replacements. The accumulated run-time on engine is relatively small compared to how long it has been installed. The replacement frequency is based upon preventive maintenance program and items that are short lived are not subject to aging management review. They are included in the preventive maintenance program at the site and time intervals adjusted, as necessary. However, the tube bundles have not been replaced and so they were not removed from LRA Table 3.3.2-24. The tubes bundle will be replaced during the PEO, and after replacement, they will no longer be subject to aging management. The replacement frequency may be reevaluated at that time commensurate with the as-found condition and industry experience.

The replacement and inspection frequency for the loss of material of the channel is based upon current operating experience and the continued, but limited, visual inspections of the heat exchanger. There have been no indications of loss of material (rusty water) discharging from the heat exchanger and no loss or gain of water in the engine side coolant nor any visible contamination. This result is based upon review of work order records since September 2015.

External inspections occur every time the engine is run.

A single industry event at North Anna on August 20, 2017 was the basis for replacement at North Anna. North Anna 1 and North Anna 2 represents the OE that is currently under

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 11 Page 3 of 4 discussion. The tube leaks were identified near the end of the initial term of the licenses. North Anna 1 and North Anna 2 entered the PEO on 4/1/2018 and 8/21/2020, respectively. This OE states a replacement frequency of 20 years, which was acceptable based upon their operating experience and not knowing when the initial leak began. The engine was likely in operation well before commercial operation, so 40 years could be assumed. A prediction was not made on how long the engine could run with the tube bundles in the as-found condition. Consequently, the plant staff declared it a Maintenance Rule functional failure, and it was later determined that criterion, zero failures to run, was too conservative. The North Anna plant staff argued it was too conservative because the diesel driven pump serves a back-up function for main motor driven pump, as it is at PNPP.

The report indicated North Anna staff did not know when the first tube began to leak. According to the OE seven tubes were found leaking. Nevertheless, the 20-year frequency at North Anna means the next replacement of the tube bundle will occur once and late into the PEO for North Anna, Unit 1.

This singular event means that other plants have not experienced this failure even though many are in their PEO. PNPPs initial 14-year frequency was chosen based upon these considerations and when there would be a convenient time to replace the tubes.

For PNPP, 2040 is 14 years after the PEO begins in 2026. However, the most convenient time to replace the heat exchanger tubes is in conjunction with the turbocharger maintenance (performed on a 6-year cycle). The last turbocharger maintenance was completed on September 17, 2021. So, the opportunities are 2027, 2033, and 2039. September 2045 exceeds the 14-year commitment. Additionally, any indication of leakage identified during annual replacement of the engine coolant means tube replacement could occur sooner without loss of the ability of the engine to carry out its mission. For example, a condition report (CR) in 2022 was generated due to low coolant level and the following is quoted from this CR. During performance of PTI-P54-P0035, Electric and Diesel Fire Pump Monthly Operability Test, the Diesel Fire Pump engine coolant level was checked per Step 5.2.2.2.c. The coolant level was found to be approximately 3 1/4" below the top of the cap, with the upper portion of the heat exchanger core uncovered. Per the NOTE in SOI-P54(WTR) Section 4.1.2, Full is approximately 2 below the top of fill cap; Heat Exchanger core shall be covered.

The work order responding to the condition determined the cause to be a leaking cap. The cap was replaced.

PNPP performs diesel driven fire pump (DDFP) full flow testing on an 18-month frequency per Functional Specifications within the Fire Protection Program. Within this test, the pump performance curve is trended against previous tests and suppression load data points from hydraulic calculations. This testing serves as another monitoring method which would identify overall pump degradation, which when it occurs, is entered into Corrective Action Program (CAP), and investigated. Reduced engine performance, due to reduction in heat exchanger effectiveness, would manifest in the results of the test and would be investigated in accordance with CAP requirements and past precedence.

When industry experience is put into perspective, one reported event out of many plants successfully running during the PEO and the relatively small amount of operating time on PNPPs engine, there is ample time to replace the tubes and no compelling reason for

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 11 Page 4 of 4 replacement before the PEO. Based upon the above history and monthly inspections of coolant level, 14 years selected is considered commensurate with the operating experience.

2. Please discuss whether the diesel driven fire pump engine monitoring and maintenance activities in PAP-1910 and PMI-0072 will continue during the period of extended operation, including before and after replacement of the coolant heat exchanger tube bundle.

Preventive maintenance activities conducted are on an annual, two-year and six-year period frequencies as required by the current site Fire Protection Program which is part of the current licensing basis (CLB) responsive to 10 CFR 50.48. The replacement activities have no bearing on these commitments. Hence, these activities will continue before and after heat exchanger replacement, consistent with the facilitys CLB.

3. Please provide sufficient information to demonstrate that the diesel driven fire pump engine monitoring and maintenance activities that will be performed by the Fire Water System program are or will be (i.e., require program enhancement(s)) equivalent to the monitoring and maintenance activities that would be performed by the Closed Treated Water Systems program.

The fire water program is a more appropriate aging management program to apply to the diesel-driven fire pump stationary drive engine than the closed treated water systems program.

Maintenance requirements are codified in the approved fire protection program per the current licensing basis and include equipment-specific requirements which are derived from manufacturers recommendations and typical industry practices. These activities will continue throughout the period of extended operation.

These maintenance requirements are specific to the diesel-driven fire pumps engine and require the engine to be drained annually, and system flushed. The diesel engine coolant and corrosion inhibitors are replaced annually, and inspections are completed to the extent practicable. These requirements, being specific to this small capacity stationary engine, are more applicable than maintenance strategies that are applied generically to large capacity cooling systems such as the plants emergency closed cooling and nuclear closed cooling systems under the closed treated water program. The engine has a coolant capacity that is very small relative to the volume of the systems within the scope of the closed treated water systems program. Closed treated water system program activities, for example, quarterly chemistry checks of the coolant, are not recommended by the manufacturer and would not increase the reliability or availability of the engine given the annual maintenance already performed under the fire protection program.

Programmatic enhancements are not needed based on plant operating experience.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 1 of 15 2 NCSG RAI-10332-R1 Question 2 Fire Water System Program Enhancements and Exceptions

Background

As stated in 10 CFR 54.29(a), one of the findings that NRC staff must make to issue a renewed license is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21.

Supplement 2 of the LRA dated June 27, 2024 (ML24180A010), stated that reconciliation of Perrys Fire Protection program with the 2011 Edition of National Fire Protection Association (NFPA) Standard 25 is underway and exceptions are expected and will be submitted in a future supplement. Supplement 2 revised the Discussion of several aging management review items in LRA Table 3.3.1 to refer to LRA Section B.2.21 for exceptions to the program; revised Table 3.3.2-24 to include plant-specific note 342 related to an exception to the inlet screens, however, there is no exception included in LRA Section B.2.21; and revised LRA Table B.1-2 to indicate that there are exceptions to the Fire Water System program.

In addition, during the breakout audit of the Fire Water System program, Perry noted that enhancements identified from the reconciliation process would be provided in a future LRA supplement.

Issue Supplement 2 of the LRA did not describe or justify an exception to the Fire Water System program related to the inlet screens. In addition, Supplement 2 of the LRA did not discuss whether additional enhancements are expected.

Request In accordance with 10 CFR 54.29(a), please identify actions that will be taken with respect to managing the effects of aging during the period of extended operation (i.e., enhancements and exceptions to the Fire Water System program) for NRC staff review.

PNPP Response In accordance with 10 CFR 54.29(a), please identify actions that will be taken with respect to managing the effects of aging during the period of extended operation (i.e.,

enhancements and exceptions to the Fire Water System program) for NRC staff review.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 2 of 15 The proposed enhancements and exceptions to the Fire Water System program follow:

Enhancements Unless otherwise noted, the following enhancements will be implemented in the identified program elements no later than six months prior to the period of extended operation.

The program will be enhanced to include inspections and testing consistent with Appendix L, Table 4a, Fire Water System Inspection and Testing Recommendations, of License Renewal Interim Staff Guidance LR-ISG-2012-02. Specific enhancements below must be considered in light of the exceptions.

General The Fire Water system program documentation will be updated by adding the enhancements required for compliance with NFPA 25 2011 Edition as either a separate attachment to the current Fire Protection Program policy administrative procedure or new plant administrative document listing the new requirements for in-scope portions of the fire water system with accompanying new Periodic Test Instructions (PTIs) to implement the testing requirements. In a few instances new Preventive Maintenance plans will be created and linked to the LR requirements.

Sprinkler Systems:

Program documents will be enhanced to require visual inspection of all in-scope sprinklers in addition to those that are directly protecting safe shutdown equipment as specified in the Fire Protection Functional Specifications. The functional specifications in the Fire Protection Program describe inspecting sprinklers in fire areas containing safe shutdown equipment on an 18-month frequency. This frequency is applied for these additional sprinklers consistent with the currently required inspection of sprinklers in fire areas containing safe shutdown equipment.

Program periodic inspection criteria will be revised to require sprinklers to be free of corrosion, foreign materials, paint and installed in the correct orientation to meet Section 5.2.1.1.1 criteria.

Program instructions will be enhanced to require inoperable sprinklers to be replaced.

These criteria include when showing signs of (1) leakage, (2) Corrosion, (3) Physical damage, (4) Loss of fluid in the glass bulb heat responsive element, (5) Loading (e.g.,

with dust), (6) Painting unless painted by the sprinkler manufacturer, or (7) any sprinkler installed incorrectly. Additionally, Annex A of NFPA 25 regarding cleaning of dust loaded sprinklers will be adapted.

The program will be enhanced to perform representative sprinkler head sampling (laboratory field service testing) or replacement of sprinkler heads within the scope of license renewal prior to exceeding the in-service (installed) limits specified in the 2011 Edition of NFPA 25. In the case of testing, requirements are selected in accordance with, and repeated at the intervals specified in, NFPA 25-2011. Testing is continued through

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 3 of 15 the period of extended operation, or until there are no untested sprinkler heads that will exceed the service limits through the remainder of the period of extended operation.

Standpipe and Hose Systems Program documentation will be revised, or new test instructions developed, to add main drains testing of the in-scope water-based standpipes including those associated with automatic water suppression systems. Program documentation will require testing of 20% of the necessary standpipe systems every refueling outage/cycle. These tests will occur every 10 years and throughout the PEO.

Private Fire Service Mains Program documentation will be revised to acknowledge compliance per Section 7.3.1.3 of NFPA 25, where underground piping supplies individual fire sprinkler, standpipe, water spray, or foam-water sprinkler systems and there are no means to conduct full flow tests, tests generating the maximum available flows shall be permitted. (Note: PNPP does not have a foam water sprinkler system.)

Program documentation will be revised to require that flow tests shall be made at flows representative of those expected during a fire, for the purpose of comparing the friction loss characteristics of the pipe with those expected for the particular type of pipe involved, with due consideration given to the age of the pipe and to the results of previous flow tests. Any flow test results that indicate deterioration of available waterflow and pressure shall be investigated to the complete satisfaction of the authority having jurisdiction to ensure that the required flow and pressure are available for fire protection.

Program documentation will be revised to include a 60-minute hydrant drainage limit requirement during testing to meet Section 7.3.2.4, NFPA 25. A note will be added to include words to the effect that due to the Plant Foundation Underdrain system, groundwater level around the nuclear island does not normally reach the level of the relevant hydrants. PNPP monitors ground water level. However, if water level were to be too high or other conditions exist to prevent drainage, the hydrant drain shall be plugged and water in the barrel shall be pumped out.

Relevant test instructions will be revised to include a statement that dry barrel hydrants that are located in areas subject to freezing weather and that have plugged drains shall be identified clearly as needing pumping after operation.

Valves and System Wide Testing See enhancement for Main Drain Testing under Standpipe and Hose Systems above.

Main Drains Testing shall require identification and correction of the cause of any 10%

reduction in full flow pressure.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 4 of 15 Water Spray Fixed Systems PNPP fire protection program documentation will be revised to require the removal, inspection for damaged and corroded parts, and cleaning of mainline strainers in water spray fixed nozzle systems in scope of License Renewal every 10 years. Adverse findings will be entered into the corrective action program for evaluation for increased frequency of inspection and trending.

In addition to flush activities currently associated with periodic flow testing, PNPP fire protection program documentation will be revised to ensure that mainline strainers are flushed after each actuation of an associated water spray fixed system.

Foam Water Systems PNPP fire protection program documentation will be revised to require that the foam liquid storage tank shall be drained of foam liquid and flushed every 10 years.

Obstruction Investigation New PNPP fire protection program documentation will be added to meet the requirements of NFPA 25, 2011 Edition, Section 14.2, Internal Inspection of Piping and Section 14.3 Obstruction Investigation and Prevention. Inspection scope established in other program elements or elsewhere in this program element, collectively referred to as existing enhancements, shall remain in effect. Where overlap or conflicts exist between existing enhancements and this enhancement: a) the existing enhancements shall take precedence, b) Section 14.2 requirements shall not apply to existing enhancements, and c) Section 14.3 guidance shall continue to apply to all inspection activities.

Program Element Affected: Detection of Aging Effects (Element 4)

As an enhancement to detect aging effects of internal surfaces of buried piping, a portion of the aboveground inspection locations will be selected where above-grade and underground or buried piping environments and material are similar, the above-grade can be extrapolated to evaluate the condition of the underground or buried piping. Program Elements Affected:

Parameters Monitored (Element 3) and Detection of Aging Effects (Element 4)

The program will be enhanced to require that when visual inspections are used to detect loss of material in the piping within the scope of license renewal, the inspection technique is capable of detecting surface irregularities that could indicate wall loss to below nominal pipe wall thickness due to corrosion and corrosion product deposition. Where such irregularities are detected, follow-up volumetric wall thickness examinations will be performed. Program Elements Affected: Parameters Monitored (Element 3) and Detection of Aging Effects (Element 4)

Enhancement i. below was implemented and identified piping configurations causing piping not to drain, hence Enhancement ii. is applicable at PNPP.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 5 of 15 i)

Prior to the period of extended operation, all accessible dry pre-action, sprinkler, horizontal pipe configurations (including fittings and pipe components) within the scope of License Renewal were walked down to provide reasonable assurance that the as-built, flow path piping system may be drained without areas that will allow water to accumulate and potentially contain corrosion products that could block the installed sprinklers. For those portions that are inaccessible, as-built drawings were used to identify such configurations:

ii)

The program will be enhanced to include augmented testing and inspections beyond those of Table 4a for portions of water-based fire protection system components within the scope of license renewal that are (a) normally dry but periodically subjected to flow (e.g., dry-pipe or pre-action sprinkler system components) and (b) cannot be drained or allow water to collect: The augmented inspections and activities are:

1. Within 5 years prior to the PEO, inspect 100% of the subject piping segment locations for trapped water and any condition such as organic and inorganic materials that might cause blockage of the sprinkler heads if the system were actuated. Any segments found to be wet or contain significant corrosion or organic matter will be cleaned and minimum wall thickness determined for the worst areas of wall loss. Results will be entered into the Corrective Action Program (CAP) for disposition and correction, as required.

2. After the completion of these inspections, monitor and record all actuations of the dry sprinkler systems within the scope of License Renewal, and

3. For any system that actuates, ensure prior to putting the dry sprinkler system back in service, that the affected system piping segments that are the subject of this issue will be inspected and any pooling water eliminated. Program Elements Affected:

Parameters Monitored (Element 3), Detection of Aging Effects (Element 4), and Operating Experience (Element 10)

Fire protection procedures will be revised, or new procedures developed to require periodic replacement of the coolant heat exchanger tube bundle on the Diesel Driven Fire Pump Engine during the period of extended operation at a frequency of every 14 years. The program will also require internal visual inspection of the heat exchanger shell and channel for loss of material in conjunction with tube bundle replacement. Program Element Affected: Detection of Aging Effects (Element 4)

The program will be enhanced to provide that if the presence of sufficient foreign organic or inorganic material to obstruct pipe or sprinklers is detected during pipe inspections, the material will be removed and its source will be determined and corrected. Program Elements Affected: Acceptance Criteria (Element 6)

The program will be augmented to perform periodic (initially, every other cycle, i.e., 4 year intervals), nonintrusive pipe thickness measurement in above ground or underground (not buried), wetted, metallic Fire Water system piping. Each 4 year sample will include at least three locations for a total of 100 feet of piping. Locations selected will be based upon system

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 6 of 15 susceptibility to corrosion, evidence of performance degradation during system flow testing or periodic flushes or prior wall thickness measurements. The method used will attempt to detect localized degradation in pipe wall thickness, e.g., Low Frequency Electromagnetic Technique (LFET), or equivalent. The idea is to use the method as a screening tool to identify "spots of interest" which are then followed up with ultrasonic (UT) testing or phased array testing (PAUT) on the spots of interest. Additionally, proximity to Safety Related or high risk equipment will be favored locations when given equivalent susceptibility or evidentiary factors. Significant findings shall be entered into the corrective action program for remediation and additional corrective actions. Significant findings will be any wall thickness less than min wall or localized minimum wall thickness more than 50% less when compared to its surroundings. Program Elements Affected: Parameters Monitored (Element 3),

Detection of Aging Effects (Element 4), and Acceptance Criteria (Element 6)

The program will be augmented for subsequent or existing leaks not yet repaired, when practical, to determine or confirm the corrosion mechanism(s) causing the leaks. The results will be processed through the Corrective Action Program to determine further actions and adjustments to the period of augmented inspections. Program Elements Affected:

Preventive Actions (Element 2)

Exceptions and Justifications are listed below:

The Fire Water System program has the following exceptions[1]:

Exception to NFPA Section Basis for the Exceptions Sprinkler Systems 5.2.1.1 Sprinklers shall be inspected from the floor level annually.

In lieu of annual inspections, PNPP will retain the current licensing basis inspection frequency. PNPP performs the sprinkler inspections once every 18 months, unless the inspection is in a high radiation area, in which case the inspection is performed every refueling cycle (24 month cycle).

As indicated in Note 5 to Table 4a of LR-ISG-2012-02, access for some inspections is feasible only during refueling outages, which are scheduled every 24 months. All sprinkler systems protecting safe shutdown equipment are in scope of license renewal; other sprinkler systems are also in scope as delineated on the scoping drawings. Inspections currently conducted in areas with safe shutdown equipment every 18 months have been satisfactory and demonstrate that more frequent inspections would not result in any different conclusion. A search of internal OE has shown very few CRs regarding obstructions and none are age related conditions but are latent or design issues.

Standpipe and Hose Systems 6.3.1.1 A flow test shall be conducted every 5 years at the hydraulically most remote hose connections of each zone of an automatic standpipe system to To flow test the hydraulically most remote hose connection of the automatic standpipe system in a manner that would provide sufficient information to verify design pressure and flow would generate a

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 7 of 15 verify the water supply still provides the design pressure at the required flow.

In lieu of adopting this code paragraph as a new requirement, PNPP will retain the current licensing basis testing requirements which demonstrate adequacy of the water supply.

Every 3 years, PNPP performs main header flow testing in the main headers that supply the standpipe system to verify that the water supply provides the largest demand design flow plus 500 gpm for hose streams over the longest route and verifies friction losses are within values used to determine design flow at design pressure based upon the measured discharge pressure and flow of a single fire pump. PNPP also performs fire pump full flow tests on an 18-month frequency.

large quantity of liquid that is potentially radwaste and could create a risk of wetting components critical to normal and shut down operations. By not performing additional flow testing, the potential for creating radwaste and increasing operational risk is reduced.

Every 3 years, PNPP partially opens hose station supply valves and confirms no flow blockage for the hoses listed in the Fire Protection Functional Specifications.

PNPP will perform main drain tests on 20 percent of the license renewal in-scope standpipes and risers requiring fire suppression functionality each refueling outage/cycle. Acceptance criteria will consist of ensuring an open flow path by verifying valve operability and flow through valve and connections with no indication of obstruction or undue restriction of water flow.

Section 6.3.1 has been revised in the 2014 Edition of NFPA 25 to indicate this testing provision is only applicable to Class I and Class III standpipe systems. PNPP interior hose stations are designed for Class II service.

6.3.1.5 A main drain test shall be performed on all standpipe systems with automatic water supplies in accordance with the requirements of Chapter 13.

In lieu of adopting this code paragraph, PNPP will adopt a sample-based testing program for main drain tests. By reference to Chapter 13 (Paragraph 13.2.5), Paragraph 6.3.1.5 invokes an annual frequency for the main drain tests. The sample-based approach will include 20% of the in-scope systems, tested on 24-month intervals.

As indicated by the note 5 in Appendix D, Table 4a of LR-ISG-2012-02, access for some inspections is feasible only during refueling outages, which are scheduled every 24 months. The number of tests to be conducted every 24 months was found to exceed the number of recommended tests or inspections (i.e., 25) in several sampling-based AMPs (e.g., XI.M38). This number of tests supports the decision that the exception still allows the plant to establish a trend if major flow blockage is occurring which is the stated purpose of this testing per the Annex A material provided in Section A.13.2.5. Main drain tests on 20 percent of the standpipes and risers every 24 months provide adequate information to determine if the condition of fire water piping is maintained consistent with design basis. Furthermore, current testing of hose stations for flow blockage every 3 years and program requirements to verify valve positions as listed below provides reasonable assurance that a trend of flow blockage would be established if it were occurring.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 8 of 15 Valve position verification:

Existing FP program includes valve position verification which addresses, in part, the considerations of a main drain test as explained in Annex A of the Code.

Fire Pumps 8.3.3.7 Suction Screens. After the waterflow portions of the annual test or fire protection system activations, the suction screens shall be inspected and cleared of any debris or obstructions.

From Annex A: A.8.3.3.7 During periods of unusual water supply conditions such as floods, inspection should be on a daily basis.

In lieu of adopting this requirement from the Code, PNPP will credit the inherent design features provided by the plant intake structure and ESW Pumphouse physical arrangement. Due to the design of the supply inlet and filtering of water from the lake to the Emergency Service Water Pumphouse Suction Bay, PNPP does not require monitoring of the suction screens on the fire pumps after the waterflow portions of the periodic tests, fire protection system activations nor during periods of unusual water supply conditions such as floods.

Requiring inspection of the fire pump suction screens in the Emergency Service Water Pumphouse Suction Bay would require diving activities at least every nine months given frequency of fire pump full flow tests. This is unnecessary given the design of the upstream inlet pipe configuration taking suction from the lake, Emergency Service Water Pumphouse traveling screens, smaller mesh size, and alternative aging management activities for the screens. The aging effects of the traveling screens is managed by the Open-Cycle Cooling Water System Program.

The design of the ESW system pump forebay incorporates traveling screens for removing submerged debris that may have entered through the intake structure. The water inlet is located more than one quarter mile offshore and submerged more than 15 feet below the surface of the lake. In order for debris to enter the ESW pumphouse, the debris would have to be submerged to the elevation of the intake structure, travel approximately 100 feet vertically downward, travel approximately 3,000 feet horizontally, and then rise vertically approximately 100 feet to the ESW pumphouse forebay. Also, the intake system is designed for an approach velocity of 0.5 fps which diminishes the uptake of debris. These features are intended to prevent any significant amount of debris from entering the ESW pumphouse forebay and clogging the travelling screens.

Water entering the suction bay travels passed a basket mesh opening of 0.375 inch in the traveling screens. If differential level from forebay to suction bay across the traveling screens exceeds 6 inches, the ESW Screen Wash pumps start, and the traveling screens start in slow speed. If the ESW Traveling Screen difference in level exceeds 10 inches, the screen shifts to fast speed. Control room alarms are provided based upon exceeding

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 9 of 15 the high level difference across the traveling screens. The size of the ESW traveling screens are intended for the ESW pump operation and therefore it is highly unlikely any significant level difference would be expected when a fire pump is operating alone.

The fire pump inlet strainer is a basket type suction strainer with a free area of at least four times the area of the suction connection and openings of such size to restrict the passage of a 1/2-inch sphere.

Since the traveling screens mesh size is smaller than the inlet strainer of the fire pumps, any debris entering from the lake would pass through the pump and therefore blockage of the inlet strainer is highly unlikely.

A search of internal condition reports in the corrective action database did not find any instances of flow blockage in fire pump suction strainers.

Valves and System-Wide Testing 13.2.5* Main Drain Test. A main drain test shall be conducted annually at each water-based fire protection system riser to determine whether there has been a change in the condition of the water supply piping and control valves. (See also 13.3.3.4.)

13.3.3.4 A main drain test shall be conducted any time the control valve is closed and reopened at system riser.

The enhancement associated with Standpipe and Hose Systems will add main drains testing to 20% of in scope water-based standpipe systems including those supplying automatic water suppression systems every refueling outage (24 months) rather than all standpipe systems annually, as required by Chapter 13 of NFPA 25.

The addition of main drain tests in enhancements under to the PNPP Fire Water Program provides compliance with this section of NFPA 25 in meeting main drains test requirement. The exception is in the number of test and to perform them annually. See the basis for exception under 6.3.1.5.

13.4.3.2.2* Each deluge valve shall be trip tested annually at full flow in warm weather and in accordance with the manufacturers instructions.

Regarding the annual testing requirement, it should be noted that the corresponding table in NUREG-2191 Volume 2, Table XI.M27-1 contains note 10, Where NFPA 25 or this table cite annual

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 10 of 15 From Annex A: A.13.4.3.2.2 *Pre-action and deluge valves in areas subject to freezing should be trip tested in the spring to allow time before the onset of cold weather for all water that has entered the system or condensation to drain to low points or back to the valve.

In lieu of the above requirement, PNPP will continue to utilize the current licensing basis testing frequencies.

PNPP conducts trip tests on open spray deluge valves in the scope of license renewal with full flow during shutdowns for refueling in accordance with 13.4.3.2.2.3. Other Deluge valve subsystems falling under sections 13.4.3.2.2.2 are tested with the supply isolation valves closed (equivalent to control valves per Chapter 3 NFPA 25) every 18 months. These systems include dry piping pre-action systems with closed sprinkler heads and ventilation filter deluge spray systems with open sprays. In the latter cases damage to equipment would occur if conducted at full or partial flow. The use of this exception is discussed below within this section and pre-action deluge valve testing is discussed below under Section 13.4.3.2.3 and 13.3.2.4.

For testing meeting the conditions under Section 13.4.3.2.2.2, PNPP trip testing of deluge valves with the isolation valve closed trapping water pressure between the isolation valve and trip valve prior to actuation. The trapped pressure causes the valve to pop open when tripped.

Any valve that fails to open is considered a failure and entered into the corrective action program to evaluate and take required corrective actions.

Considering the above, the exception is not with the method of trip testing every testing or inspections, testing and inspections can be conducted on a refueling outage interval if plant-specific OE has shown no loss of intended function of the in-scope systems due to aging effects being managed for the specific component (e.g., loss of material, flow blockage due to fouling). Furthermore, as noted, NFPA 25 section 13.4.3.2.2.3 permits waiting when testing is possible without risking equipment damage and plant transients due to energized electrical components. Although partial blockage occurs in outdoor open spray deluge testing of transformers, the spray patterns have been found acceptable.

Each time partial blockage is identified the blockages are cleared. Thus, plant specific OE supports continued testing at a frequency of each refueling outage.

Hydrogen seal oil deluge valve open spray system:

PNPP proposes to maintain the 5-year interval for full flow testing of the Hydrogen seal oil deluge system. Hydrogen seal oil spray nozzle and piping are located completely with an enclosed structure, i.e. the Turbine Building. Since the piping and nozzles are indoors, these nozzles are not subject to freezing, outdoor cycles, moisture, nor biological intrusion. This makes plugging of nozzles unlikely.

During the full flow test every 5 years, water flow is captured and processed within the liquid radwaste systems. Due to having only 8 nozzles it is estimated that at least 500 gallons of radwaste would be generated. The hydrogen seal oil deluge valve trip test is conducted at a frequency of every 12 months. To further support this exception, the history of Hydrogen Seal Oil full flow spray testing since 2007 through 2021 there has been no identified plugged or partially plugged nozzles. In contrast and to further support this exception, the difference between indoor hydrogen seal oil testing and outdoor testing of open spray nozzles is illustrated as follows. In 2019, a condition report was written documenting 97 examples of outdoor, open spray nozzles found to have some level of partially plugging in a four-year period from 2016 thru 2019. The affected nozzles were found to have an acceptable spray pattern and were subsequent cleaned. As noted above, over the past 17 years no instances of plugging of indoor

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 11 of 15 18 months but with the requirement for annual testing.

Further, full flow spray subsystem protecting Hydrogen Seal Oil system equipment are conducted every 5 years and trip tests with the supply isolation valve closed (equivalent control valve closed) are conducted annually.

NFPA 25, 13.4.3.2.2.2 states: Where the nature of the protected property is such that water cannot be discharged for test purposes, the trip test shall be conducted in a manner that does not necessitate discharge in the protected area. The PNPP deluge valves with spray systems protecting ventilation filter / charcoal plenums test deluge valves perform trip tests with isolation valves closed as described above to protect the equipment that would be damaged by water spray. Additionally, these tests verify that the ventilation plenum drain valves open upon activation of the spray system. PNPP deluge valves with spray systems protecting ventilation filter / charcoal plenums are discussed further in the exception for 13.4.3.2.2.5 below.

NFPA 25, 13.4.3.2.2.3, where the nature of the protected property is such that water cannot be discharged unless protected equipment is shut down (e.g.,

energized electrical equipment), a full flow system test shall be conducted at the next scheduled shutdown.

Deluge water open spray systems protecting ventilation filter plenums and closed sprinkler head pre-action systems that protect sensitive equipment in the cable spreading rooms and main turbine bearings meet this requirement. An exception for these pre-action systems is discussed further under 13.4.3.2.3 and 13.4.3.2.4 Hydrogen Seal Oil open spray nozzles were noted in the work order notes and a search of OE found no condition reports regarding plugging of the HSO nozzles.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 12 of 15 13.4.3.2.2.5 The water discharge patterns from all of the open spray nozzles or sprinklers shall be observed to ensure that patterns are not impeded by plugged nozzles, that nozzles are correctly positioned, and that obstructions do not prevent discharge patterns from wetting surfaces to be protected.

13.4.3.2.2.5 (A) Where the nature of the protected property is such that water cannot be discharged, the nozzles or open sprinklers shall be inspected for correct orientation and the system tested with air to ensure that the nozzles are not obstructed.

In lieu of the above, PNPP will continue to utilize the testing approved in the plants CLB. Ventilation filter unit plenums in scope cannot be tested with water and have no provisions to perform an air test to verify that the spray openings are not obstructed.

Ventilation filter unit plenums in scope cannot be tested with water and have no provisions to perform an air test to verify that the spray openings are not obstructed. The Fire Protection Program Functional Specifications requires that each charcoal filter plenum spray header/nozzle is visually inspected each time the charcoal is changed. This activity will ensure that there are no debris locally that would obstruct the spray nozzles if the spray system were actuated.

13.4.3.2.3 Except for pre-action systems covered by 13.4.3.2.5, every 3 years the pre-action valve shall be trip tested with the control valve fully open.

13.4.3.2.4 During those years when full flow testing in accordance with 13.4.3.2.3 is not required, the pre-action valve shall be trip tested with the control valve partially open.

In lieu of the above requirements from the Code, PNPP will continue to utilize the testing frequency and scope prescribed in the plants current license basis. PNPP pre-action systems are tested with the isolation valves closed (Equivalent to control valve per NFPA 25 definitions Chapter 3) every 18 months.

The method of trip testing of the pre-action deluge valves is the same as the method discussed for the deluge valves Pre-action system:

Trip testing the pre-action valves with the control valve fully open would allow fire water to enter the normally dry portion of the system in the case of dry pipe systems which is trying to be avoided. In addition, there is a potential for wetting equipment critical to normal and shut down operations if one of the closed sprinkler heads actuate.

PNPP proposes to maintain the current testing intervals on the basis that PNPP has demonstrated adequate performance and that adopting a more frequent testing regimen is not expected to improve system performance or availability but adds to the risk. Furthermore, The corresponding table in NUREG-2191 Volume 2, Table XI.M27-1 contains note 10 Where NFPA 25 or this table cite annual testing or inspections, testing and inspections can be conducted on a refueling outage interval if plant-specific OE has shown no loss of intended function of the in-scope systems due to aging effects being managed for the specific component (e.g., loss of material, flow blockage

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 13 of 15 serving open spray systems and is not repeated here. The pre-action system testing falls under allowances described in paragraphs 13.4.3.2.2.2 and are not considered exceptions, but for clarity, the basis for exception expands upon this.

due to fouling). PNPP meets this condition by testing at the current frequency.

To ensure these systems will perform the intended function throughout the PEO, PNPP proposes trip testing and inspections that ensure the downstream piping is free of matter that might plug system sprinklers if the system were actuated. As noted in enhancement ii, augmented inspections of portions of water-based fire protection system components within the scope of license renewal that are (a) normally dry but periodically subjected to flow (e.g., dry-pipe or pre-action sprinkler system components) and (b) cannot be drained or allow water to collect, three (3) activities will be performed. These activities provide a baseline for ensuring that the piping does not contain inorganic or organic materials that could plug downstream sprinklers upon a system actuation. Further, they include continual monitoring of the system for actuation, and if actuated, restores the piping to the baseline condition before restoring the system to service. Collectively this enhancement ensures the piping system is clear and the new enhancement for performing main drains testing ensures the in-scope systems can perform its intended function throughout the PEO. The in-scope pre-action sprinkler systems are the Unit 1 and 2 Cable Spreading Divisions 1 and 2 systems and the Unit 1 Main Turbine Driven Pump Bearings Pre-action systems. The downstream piping in the Unit 1 Main Turbine Driven Pump Bearings Pre-action systems is dismantled periodically during refueling outages in conjunction with main turbine maintenance. In essence, if corrosion or inorganic materials would be discovered this condition would be entered into the corrective action program to evaluate and take corrective actions.

Water Spray Fixed Systems 10.2.1.7 Mainline strainers shall be removed and inspected every 5 years for damaged and corroded parts.

In lieu of the above five-year requirement, PNPP will perform this activity on a ten-year frequency. PNPP fire protection program inspects mainline supply strainers for damage including corrosion and flow blockage PNPP internal operating experience demonstrates that the current frequency of inspection is adequate to support fixed water system sprays.

Except for testing the Hydrogen Seal Oil fixed water spray systems every 5 years, the other spray systems in scope of License Renewal are tested every refueling cycle.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 14 of 15 from organic and inorganic debris in water spray fixed nozzle systems in scope of License Renewal every 10 years. Adverse findings are entered into the corrective action program for evaluation for increased frequency of inspection and for trending.

10.2.7* Strainers.

10.2.7.1 Mainline strainers (basket or screen) shall be flushed until clear after each operation or flow test.

From Annex A: A.10.2.7 *Mainline strainers should be removed and inspected for damaged and corroded parts every 5 years.

Due to system design limitations, PNPP will not explicitly confirm the flushed until clear portion of this requirement.

PNPP full flow tests of each fixed open spray system includes steps to flush the associated main line strainer after flow testing of all associated fixed spray system tests are completed. These tests are performed every refueling outage.

Guidance is provided to limit flush water flowrate to prevent backup of floor drains. During flush operation, the strainer handwheel is rotated a full 2 turns. Flush water can be viewed through sight glass provided in the flush connection. Due to limitations on ability to confirm flush water is clear, flush water clarity is not included in guidance.

See response to exception in 10.2.1.7 above.

Strainer flush lines are connected to the floor drain system piping and discharge cannot be viewed directly. However, Flush water can be viewed through sight glass provided in the flush connection. Floor drains are susceptible to backing up when subjected to high in-flow rates. The testing recommends that personnel be staged on each floor between the flushing elevation and the sump pit elevation to monitor for backup. Water back-up out of the floor drains shall be treated as potentially contaminated. Guidance encourages limiting flush time to as minimal as possible.

Nevertheless, strainers are cycled two turns during flush operation to ensure debris removal.

Operating experience has demonstrated that these activities have been successful in maintaining strainers without having any significant blockage.

Consequently, due to system design limitations, PNPP will not explicitly confirm the flushed until clear portion of this requirement but has presented a successful alternative.

Notes:

1.

Exceptions are not taken when the equipment discussed in the requirement does not exist at PNPP.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 12 Page 15 of 15 The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References NFPA 25 2011 Edition Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 13 Page 1 of 5 3 NCSG RAI-10337-R1 Regulatory Basis Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S.

Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described in the requests for information.

Question 1 (Drywell Mechanical Penetrations - Aging Management Programs)

Background

License Renewal Application (LRA) Supplement 3 (ML24206A150) updated the discussion of Aging Management Review (AMR) item 3.5.1-10 in LRA Table 3.5.1 to state, in part, SCC

[stress corrosion cracking] is managed by ASME [American Society of Mechanical Engineers]

Section XI, Subsection IWE and Structures Monitoring and Fire Protection programs for Drywell mechanical penetrations. See Further evaluation section 3.5.2.2.1.6. LRA Supplement 3 also revised LRA Section 3.5.2.2.1.6 to state, [a]lthough an aggressive chemical environment doesnt exist, the potential for SCC is assumed for these components, and the aging effect is managed by the ASME Section XI, Subsection IWE and 10 CFR 50, Appendix J programs for containment penetrations, and by the ASME Section XI, Subsection IWE, Fire Protection and Structures Monitoring programs for drywell mechanical penetrations.

Issue LRA Supplement 3 revised LRA Table 3.5.2-1 to cite AMR item 3.5.1-10 for managing cracking of stainless steel drywell mechanical penetrations by deleting the Structures Monitoring program and adding the 10 CFR 50, Appendix J program. This change is inconsistent with the statements made for both AMR item 3.5.1-10 in LRA Table 3.5.1 and LRA Section 3.5.2.2.1.6, where the Structures Monitoring program and not the 10 CFR 50, Appendix J program is credited for managing SCC of stainless steel drywell mechanical penetrations.

Request Please reconcile the discrepancies in the programs credited for managing cracking of stainless steel drywell mechanical penetrations in AMR item 3.5.1-10 of LRA Table 3.5.1, LRA Section 3.5.2.2.1.6, and LRA Table 3.5.2-1.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 13 Page 2 of 5 PNPP Response Please reconcile the discrepancies in the programs credited for managing cracking of stainless steel drywell mechanical penetrations in AMR item 3.5.1-10 of LRA Table 3.5.1, LRA Section 3.5.2.2.1.6, and LRA Table 3.5.2-1.

PNPPs review of this issue has concluded that the current text in LRA Table 3.5.1 item 10 and LRA Section 3.5.2.2.1.6 are both consistent and are accurate. The changes made to the rows in Table 3.5.2-1 for the component drywell mechanical penetrations are not consistent with LRA Table 3.5.1 item 10 and LRA Section 3.5.2.2.1.6, since the 10 CFR50 Appendix J Program does not apply to drywell leak rate testing and the drywell mechanical penetrations.

Accordingly, the following change is proposed for LRA Table 3.5.2-1 to restore consistency with LRA Table 3.5.1 item 10 and LRA Section 3.5.2.2.1.6:

Row 96 Component type: Drywell mechanical penetrations Intended Function: EN, SPB, SSR, FB, SRE Material: Stainless steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking AMP: ASME Section XI, Subsection IWE and Structures monitoring NUREG 1801 Item: II.B4.CP-38 Table 1 Item: 3.5.1-10 Notes: A Additionally, upon further review, it was determined that the use of 10 CFR Part 50, Appendix J aging management program does not apply to the following components in the drywell and therefore, the following rows in Table 3.5.2-1 will be revised to replace the 10 CFR Part 50, Appendix J AMP with the Structures Monitoring AMP:

Table 3.5.2-1:

Component Type: Drywell electrical penetrations Intended Function: EN, SPB, SSR, FB, SRE Material: Steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of material AMP: ASME Section XI, Subsection IWE and Structures Monitoring NUREG-1801 Item: II.B3.1.CP-43 Table 1 Item: 3.5.1-35 Note: E

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 13 Page 3 of 5 Component Type: Drywell equipment hatch Intended Function: EN, MB, SPB, SSR Material: Steel Environment: Air - indoor, uncontrolled (Int)

AERM: Loss of material AMP: ASME Section XI, Subsection IWE and Structures Monitoring NUREG-1801 Item: II.B3.1.CP-43 Table 1 Item: 3.5.1-35 Note: E Component Type: Drywell Equipment Hatch Seals Intended Function: EN, MB, SPB, SSR Material: Elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Change in material properties and cracking AMP: Structures Monitoring NUREG-1801 Item: II.B4.CP-41 Table 1 Item: 3.5.1-33 Note: E Component Type: Drywell Equipment Hatch Seals Intended Function: EN, MB, SPB, SSR Material: Elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking AMP: Structures Monitoring NUREG-1801 Item: II.B4.CP-41 Table 1 Item: 3.5.1-33 Note: E Component Type: Drywell Equipment Hatch Seals Intended Function: EN, MB, SPB, SSR Material: Elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of sealing AMP: Structures Monitoring NUREG-1801 Item: II.B4.CP-41 Table 1 Item: 3.5.1-33 Note: E Component Type: Drywell head Intended Function: EN, MB, SPB, SSR Material: Steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of material AMP: ASME Section XI, Subsection IWE and Structures Monitoring NUREG-1801 Item: II.B3.1.CP-43 Table 1 Item: 3.5.1-35 Note: E

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 13 Page 4 of 5 Component Type: Drywell liner plate Intended Function: SPB, SSR Material: Steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of material AMP: ASME Section XI, Subsection IWE and Structures Monitoring NUREG-1801 Item: II.B3.1.CP-43 Table 1 Item: 3.5.1-35 Note: E Component Type: Drywell personnel airlock Intended Function: EN, SPB, SSR Material: Steel Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of material AMP: ASME Section XI, Subsection IWE and Structures Monitoring NUREG-1801 Item: II.B3.1.CP-43 Table 1 Item: 3.5.1-35 Note: E Component Type: Drywell personnel airlock Intended Function: EN, SPB, SSR Material: Steel Environment: Air - indoor, uncontrolled (Int)

AERM: Loss of material AMP: ASME Section XI, Subsection IWE and Structures Monitoring NUREG-1801 Item: II.B3.1.CP-43 Table 1 Item: 3.5.1-35 Note: E Table 3.5.1, Item 3.5.1-35 will be revised to state as follows:

From:

Consistent with NUREG-1801. The ASME Section XI, Subsection IWE and 10 CFR 50, Appendix J Programs will manage loss of material of steel elements associated with the containment pressure boundary.

To:

Consistent with NUREG-1801 with the following clarification. The ASME Section XI, Subsection IWE and 10 CFR 50, Appendix J Programs will manage loss of material of steel elements associated with the containment pressure boundary. Loss of material in drywell steel components will be managed by ASME Section XI, Subsection IWE and Structures Monitoring Programs.

Table 3.5.1, Item 3.5.1-33 will be revised to state as follows:

From:

Consistent with NUREG-1801, with a different program assigned for some components and with the following clarifications. Aging of elastomer seals will be managed by the 10 CFR 50, Appendix J Program. The Structures Monitoring Program will manage aging

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 13 Page 5 of 5 of elastomer upper containment pool gate seals which are not within the scope of the 10 CFR 50, Appendix J Program although located within the Containment Structure. Fire Protection Program will manage the loss of sealing in the Shield Building electrical penetration seals and sealants having a fire barrier (FB) intended function. Change in material properties and cracking is aligned with the aging effect loss of sealing. The External Surfaces Monitoring of Mechanical Components will manage change in material properties and cracking of elastomer seals and gaskets (door, manway, and hatch) exposed to Air-indoor, uncontrolled in Bulk Civil Commodities. Structures Monitoring Program will manage change in material properties (hardening and loss of strength) and cracking in elastomeric components exposed to outdoor weather. Other elastomer seals and gaskets and fire penetration seals exposed to air are addressed under items 3.3.1-57 and 3.5.1-72.

To:

Consistent with NUREG-1801, with a different program assigned for some components and with the following clarifications. Aging of containment elastomer seals will be managed by the 10 CFR 50, Appendix J Program. Aging of elastomer Drywell Equipment Hatch Seals will be managed by Structures Monitoring Program. The Structures Monitoring Program will manage aging of elastomer upper containment pool gate seals which are not within the scope of the 10 CFR 50, Appendix J Program although located within the Containment Structure. The Fire Protection Program will manage the loss of sealing in the Shield Building electrical penetration seals and sealants having a fire barrier (FB) intended function. Change in material properties and cracking is aligned with the aging effect loss of sealing. The External Surfaces Monitoring of Mechanical Components will manage change in material properties and cracking of elastomer seals and gaskets (door, manway, and hatch) exposed to Air-indoor, uncontrolled in Bulk Civil Commodities. The Structures Monitoring Program will manage change in material properties (hardening and loss of strength) and cracking in elastomeric components exposed to outdoor weather. Other elastomer seals and gaskets and fire penetration seals exposed to air are addressed under items 3.3.1-57 and 3.5.1-72.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 14 Page 1 of 3 4 NCSG RAI-10337-R1 Question 2 (Pyrocrete)

Background

LRA Supplement 2 (ML24180A010) revised LRA Sections A.1.20 and B.2.20 to state that the Fire Protection program manages loss of material, cracking/ delamination, and separation of Pyrocrete fireproofing. LRA Supplement 3 (ML24206A150) revised LRA Table 3.5.2-4 to credit the Fire Protection program with managing separation of Pyrocrete fireproofing (cracking and loss of material originally included). Delamination was not added to LRA Table 3.5.2-4. During the audit of the Fire Protection program, the NRC staff noted that Revision 6 of LRPY-CAMR-001, Structural Material/Environment/Aging Effect Bases Report, states that change in material properties due to chemical exposure is not applicable for Pyrocrete because if it were exposed to corrosive chemicals, it would be event-driven and not age related. In addition, the report states that change in material properties due to gamma irradiation exposure is not applicable for Pyrocrete because Pyrocrete is a cementitious material and not an organic polymer, such that it is less susceptible to irradiation aging effects.

Issue Delamination was identified as an applicable aging effect for Pyrocrete in LRA Supplement 2 (LRA Sections A.1.20 and B.2.20), however, it was not added to LRA Table 3.5.2-4 in subsequent LRA supplements. Therefore, based on the LRA supplements and Revision 6 of LRPY-CAMR-001, it is unclear whether the applicant considers delamination applicable or not applicable for Pyrocrete.

Section 6.3.3.2 of EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), November 2018, states that 106 rads, which is the organic polymer radiation damage threshold, is conservative for fire wraps and fire stops, and if the radiation dose is below 106 rads, then change in material properties is not expected. Table 6-3 of EPRI 3002013084 includes change in material properties due to gamma irradiation exposure for cementitious fireproofing as being applicable for exposures exceeding 106 rads. Given that Revision 6 of LRPY-CAMR-001 only states Pyrocrete is less susceptible to irradiation aging effects, it is unclear why change in material properties due to gamma irradiation exposure was not identified as applicable.

Request

1. Please discuss whether delamination is an applicable aging effect for Pyrocrete.

2. Please provide additional justification for why change in material properties due to gamma irradiation exposure is not an applicable aging effect for Pyrocrete (e.g., not located in areas where the radiation dose would exceed 106 rads).

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 14 Page 2 of 3 PNPP Response

1. Please discuss whether delamination is an applicable aging effect for Pyrocrete.

Per EPRI Structural Tools Table 6-2, the aging effect Cracking/Delamination is an applicable aging effect for fire wraps subjected to vibration. Pyrocrete is a fire wrap material per EPRI Structural Tools Section 6.1. LRA Table 3.5.2-4 identified only Cracking as an aging effect for pyrocrete. Therefore, the aging effect for pyrocrete in LRA Table 3.5.2-4 will be changed to Cracking/Delamination.

2. Please provide additional justification for why change in material properties due to gamma irradiation exposure is not an applicable aging effect for Pyrocrete (e.g., not located in areas where the radiation dose would exceed 106 rads).

EPRI Structural Tools Section 6.3.3 states as follows:

6.3.3.2 Irradiation Damage Table 6-3 of EPRI 3002013084 indicates that change in material properties due to gamma irradiation exposure for cementitious fireproofing is applicable for exposures exceeding 1 E 6 rads.

According to PNPP Specification SP-2157 R2, Technical Requirements for Procurement and Installation of Pyrocrete 241 Fire Proofing, at PNPP, installation of pyrocrete is limited to the following applications/locations:

Structural steel columns and beams in the Control Complex, Fuel oil day tanks and supports in the Diesel Generator Building, and Fire dampers throughout Unit 1 Of these applications, only pyrocrete applied to fire damper housings would be located in areas where integrated radiation dose over 60 years of operation may exceed 1E 6 rads.

Therefore, it is proposed that a new bulk component of fireproofing, fire damper housing will be added to LRA Table 3.5.2-4, including a new plant specific note as follows:

Component type: Fireproofing, fire damper housing Intended Function: FB, SRE Material: Pyrocrete Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking/delamination AMP: Fire Protection NUREG 1801 Item: N/A Table 1 Item: N/A Notes: H 505

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 14 Page 3 of 3 Component type: Fireproofing, fire damper housing Intended Function: FB, SRE Material: Pyrocrete Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of material AMP: Fire Protection NUREG 1801 Item: N/A Table 1 Item: N/A Notes: H 505 Component type: Fireproofing, fire damper housing Intended Function: FB, SRE Material: Pyrocrete Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG 1801 Item: N/A Table 1 Item: N/A Notes: H 505 Component type: Fireproofing, fire damper housing Intended Function: FB, SRE Material: Pyrocrete Environment: Air - indoor, uncontrolled (Ext)

AERM: Change in material properties, cracking AMP: Fire Protection NUREG 1801 Item: N/A Table 1 Item: N/A Notes: H 540 Note 540 - The fire protection program manages this aging effect. This aging affect only applies to pyrocrete located where integrated dose will exceed 1 E 6 rads, in the following environmental zones: AB-7; AB-8; AB-10; FB-6; TB-1 The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 1 of 9 5 NCSG RAI-10337-R1 Question 3 (Fiberglass/Alumina Silicate/Calcium Silicate/Mineral Fiber)

Background

LRA Supplement 3 (ML24206A150) revised LRA Table 3.5.2-1 to credit the Fire Protection program with managing cracking/delamination and loss of material of fiberglass/alumina silicate/calcium silicate/mineral fiber drywell mechanical penetrations (fiberglass). LRA Supplement 3 revised LRA Table 3.5.2-4 to credit the Fire Protection program with managing cracking delamination and loss of material of fiberglass/alumina silicate/calcium silicate/mineral fiber fire wrap. Also, LRA Supplement 3 revised LRA Table 3.5.2-4 to credit the Fire Protection program with managing cracking/delamination, loss of material, and separation of fiberglass/alumina silicate/calcium silicate/mineral fiber penetration sealant (fire).

Table 3.5.2-4 in the initial LRA cited no aging effects requiring management and no aging manage program for fiberglass/alumina silicate/calcium silicate/mineral fiber insulation and penetration sealant (flood, radiation) exposed externally to uncontrolled indoor air, and insulation exposed externally to outdoor air. Plant-specific note 503 was cited and states, Operating experience review did not identify aging effects that affect intended function for these material/environment combinations. These components do not have a fire barrier intended function.

Issue It is unclear why separation was cited as an applicable aging effect for the fiberglass/alumina silicate/calcium silicate/mineral fiber penetration sealant (fire) but not cited for the fiberglass/alumina silicate/calcium silicate/mineral fiber drywell mechanical penetrations (fiberglass). Penetrations would appear to be associated with the separation of fire zones. In addition, it is unclear why change in material properties was not cited as an applicable aging effect for fiberglass/alumina silicate/calcium silicate/mineral fiber drywell mechanical penetrations (fiberglass), fire wrap, and penetration sealant (fire).

For fiberglass/alumina silicate/calcium silicate/mineral fiber, Revision 6 of LRPY-CAMR-001 appears to only evaluate a loss of insulation function and not a fire barrier intended function, and references Table 10-7 of EPRI 3002013084. The staff notes that Section 10.3.1 of EPRI 3002013084 states that [i]nsulating materials associated with fire barriers are addressed in Section 6 of this report. Therefore, based on Table 6-2 of EPRI 3002013084, separation due to vibration, movement, and shrinkage may be applicable to fire stops, and change in material properties due to gamma irradiation exposure may be applicable to fire wraps and fire stops.

While LRA Supplement 2 (ML24180A010) added separation in elastomer and pyrocrete, to LRA Sections A.1.20 or B.2.20, subsequent LRA supplements have not revised LRA Sections A.1.20 or B.2.20 to state that the Fire Protection program will manage separation of fiberglass/alumina silicate/calcium silicate/mineral fiber as indicated in LRA Table 3.5.2-4 for fiberglass/alumina silicate/calcium silicate/mineral fiber penetration sealant (fire).

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 2 of 9 EPRI 3002013084 notes that insulation systems typically consist of the insulating material and a barrier or covering that provides protection for the specific application. It is unclear if all the fiberglass/alumina silicate/calcium silicate/mineral fiber component types (fire barriers and non-fire barriers) installed at Perry have a protective barrier or covering. In addition, during the audit of the Fire Protection program it was stated that insulation material in high gamma radiation areas is limited by design to metallic insulation. It is unclear whether the non-fire barrier fiberglass/alumina silicate/calcium silicate/mineral fiber component types installed at Perry are located in areas of high gamma radiation.

Request 1.

Please discuss why separation was cited as an applicable aging effect for the fiberglass/alumina silicate/calcium silicate/mineral fiber penetration sealant (fire) but not cited for the fiberglass/alumina silicate/calcium silicate/mineral fiber drywell mechanical penetrations (fiberglass).

2.

Please discuss why change in material properties was not cited as an applicable aging effect for fiberglass/alumina silicate/calcium silicate/mineral fiber drywell mechanical penetrations (fiberglass), fire wrap, and penetration sealant (fire) (e.g.,

not located in areas where the radiation dose would exceed 10 6 rads).

3.

Please discuss why LRA Sections A.1.20 and B.2.20 have not been revised to state that the Fire Protection program will manage separation of fiberglass/alumina silicate/calcium silicate/mineral fiber.

4.

Please discuss whether all the fiberglass/alumina silicate/calcium silicate/mineral fiber component types (fire barriers and non-fire barriers) installed at Perry have a protective barrier or covering as described in EPRI 3002013084.

5.

Please discuss whether any of the non-fire barrier fiberglass/alumina silicate/ calcium silicate/mineral fiber component types installed at Perry are located in areas of high gamma radiation. If they are, then discuss whether change in material properties is an applicable aging effect.

PNPP Response Request 1.

Please discuss why separation was cited as an applicable aging effect for the fiberglass/alumina silicate/calcium silicate/mineral fiber penetration sealant (fire) but not cited for the fiberglass/alumina silicate/calcium silicate/mineral fiber drywell mechanical penetrations (fiberglass).

PNPP agrees that separation is an applicable aging effect for drywell mechanical penetrations.

Per EPRI Structural Tools, EPRI report 1015078 and EPRI 3002013084, Section 6.3.4, Separation is the destruction of the adhesion between a fire barrier material and an adjacent surface.

Drywell mechanical penetrations represent a fire barrier between fire zones 1RB-1b and RB-1c as reflected in UFSAR Figure 9A-2, Figure 9A-5, Figure 9A-10, Figure 9A-18, and Figure 9A-22.

As cited in LRA Table 3.5.2-1, the component Drywell Mechanical Penetration (fiberglass) includes the material fiberglass/alumina silicate/calcium silicate/mineral fiber. Based on the fire

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 3 of 9 barrier intended function, separation is concluded to be an applicable aging effect for this application of fiberglass/alumina silicate/calcium silicate/mineral fiber.

The following LRA changes are proposed to address this aging effect:

LRA Table 3.5.2-1 will have a new row added as follows:

Component Type: Drywell mechanical penetration (fiberglass)

Intended Function: FB, SRE Material: Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 In LRA Sections A.1.20 and B.2.20 for the list of aging effects and materials in the program description, update the last bullet as follows:

From:

Separation in elastomer and pyrocrete To:

Separation in elastomer, pyrocrete, and fiberglass/alumina silicate/calcium silicate/mineral fiber 2.

Please discuss why change in material properties was not cited as an applicable aging effect for fiberglass/alumina silicate/calcium silicate/mineral fiber drywell mechanical penetrations (fiberglass), fire wrap, and penetration sealant (fire) (e.g., not located in areas where the radiation dose would exceed 10 6 rads).

Section 6.3.3.2 in of EPRI report 1015078 (Structural Tools) and EPRI 3002013084 (Structural tools for SLR) state the same information, i.e. The reported radiation damage threshold value for organic polymers of 10^6 rads is conservative for fire wrap and fire stop materials (Bisco SF-20 silicon foam, for example, resists radiation effects up to 2 x 10^8 rads) and may not affect the fire retarding properties of fire wrap or fire stop materials at these levels.

Table 6-2, Applicable Aging Effects for Fire Wrap and Fire Stops, in EPRI report 1015078 and EPRI 3002013084, indicate the applicability of change in material properties is at a threshold above 10^6 rads and it is Not Applicable below that threshold. The Fire Protection Program has robust inspection criteria and will identify change in material properties. Therefore, PNPP will add change in material properties for these materials and components with standard Note H and a plant specific note explaining it is applicable to areas where the integrated dose is above the threshold value 10^6 rads.

Among the intended functions listed, these components also have a fire barrier function.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 4 of 9 Proposed LRA changes:

LRA Table 3.5.2-1 and Table 3.5.2-4 will have a new row(s) added as follows:

Table 3.5.2-1:

Component Type: Drywell mechanical penetration (fiberglass)

Intended Function: FB, SRE Material: Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber Environment: Air - indoor, uncontrolled (Ext)

AERM: Change in material properties AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-4 Component Type: Fire wrap Intended Function: FB, SRE Material: Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber Environment: Air - indoor, uncontrolled (Ext)

AERM: Change in material properties AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-4 Component Type: Penetration sealant (fire)

Intended Function: FB, SRE Material: Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber Environment: Air - indoor, uncontrolled (Ext)

AERM: Change in material properties AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 In Sections A.1.20 and B.2.20 for the list of aging effects and materials in the program description, replace the following bulleted paragraph with respect to reduced thermal resistance from physical damage / degradation:

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 5 of 9 From:

Change in material properties of concrete (loss of bond and reduction of strength),

elastomer (such as cracking or crazing, swelling, discoloration and melting), various types fireproofing, seals & sealants, fire wrap materials 3M Interam, Unimpregnated and impregnated (with elastomer) fiber glass fabric cracking, delaminating and visible deterioration)

To:

Change in material properties of concrete (loss of bond and reduction of strength),

elastomer (such as cracking or crazing, swelling, discoloration and melting), various types of fireproofing, seals & sealants, fire wrap materials 3M Interam, Unimpregnated and impregnated (with elastomer) fiber glass fabric (cracking, delaminating and visible deterioration), and Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber (Reduced thermal resistance / degradation from gamma irradiation threshold above 10^6 rads).

3.

Please discuss why LRA Sections A.1.20 and B.2.20 have not been revised to state that the Fire Protection program will manage separation of fiberglass/alumina silicate/calcium silicate/mineral fiber.

LRA Sections A.1.20 and B.2.20 should have stated the Fire Protection Program will manage separation of fiberglass/alumina silicate/calcium silicate/mineral fiber.

See changes to LRA in Question 3, Request 1.

4.

Please discuss whether all the fiberglass/alumina silicate/calcium silicate/mineral fiber component types (fire barriers and non-fire barriers) installed at Perry have a protective barrier or covering as described in EPRI 3002013084.

Non-Fire Barrier Intended Function From a review of the PNPP specification for insulation of plant systems, thermal insulation, including materials such as fiberglass, alumina silicate, calcium silicate, and mineral fiber is categorized into classes. All thermal insulation classes are jacketed, though some classes use different jacketing. Thus, in-scope insulation component types comprised of fiberglass, alumina silicate, calcium silicate, and mineral fiber are considered to include a protective barrier or covering as described in EPRI 3002013084.

EPRI 3002013084 Section 10.3.3 Aging Effects Mechanical damage (for example, damage from personnel stepping on the insulation or impacting it during maintenance) of the protective barriers and/or coverings of insulation are not considered to be a passive aging mechanism.

However, Table 10-7 in the same EPRI document includes an evaluation of the aging effect reduced thermal insulation resistance / moisture intrusion material for the material Nonmetallic insulation material - calcium silicate, asbestos, perlite, ceramic fiber, mineral fiber, fiberglass, ceramic, woven glass fiber that states:

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 6 of 9 Mechanism Applicability Criteria Although protective coverings protect the insulation materials from change in material properties, degradation and damage of the protective coverings is possible; therefore, degradation of insulating materials is an aging effect that requires management in both air-indoor and air-outdoor environments unless plant-specific operating experience can justify that no damage to the insulation protective coverings has occurred.

The aging effect is managed by External Surfaces Monitoring of Mechanical Components Program, where damaged protective covers would be identified and entered in the Corrective Action Program.

The above discussion results in adding change in material properties as an applicable aging mechanism for both fire barrier and non-fire barrier component types composed of fiberglass/alumina silicate/calcium silicate/mineral fiber.

LRA Changes:

LRA Changes are required for LRA Table 3.5.2-4, Bulk Commodities, for the component type Insulation with material and environments Fiberglass/ Alumina silicate/ Calcium silicate/

Mineral fiber subjected to Air-indoor, uncontrolled and Air-outdoor, which currently states no aging effects. These rows will be replaced with the following:

Table 3.5.2-4 Component Type: Insulation Intended Function: SNS Material: Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber Environment: Air - indoor, uncontrolled (Ext)

AERM: Change in material properties AMP: External Surfaces Monitoring of Mechanical Components NUREG-1801 Item: VIII.I.S-403 (LR-ISG-2012-02)

Table 1 Item: 3.4.1-64 (LR-ISG-2012-02)

Note: A Table 3.5.2-4 Component Type: Insulation Intended Function: SNS Material: Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber Environment: Air - Outdoor (Ext)

AERM: Change in material properties AMP: External Surfaces Monitoring of Mechanical Components NUREG-1801 Item: VIII.I.S-403 (LR-ISG-2012-02)

Table 1 Item: 3.4.1-64 (LR-ISG-2012-02)

Note: A

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 7 of 9 Revise Table 1 Item Discussion 3.4.1-64 (LR-ISG-2012-02) as follows:

From:

Not Applicable. This item is not used in the Steam and Power Conversion Systems.

There are no Jacketed calcium silicate or fiberglass insulation subject to aging management with an intended function of thermal insulation in the Bulk Civil Commodities.

To:

Consistent with NUREG-1801 with the following clarification. This item is not used in the Steam and Power Conversion Systems. In the Bulk Civil Commodities aging management review, the External Surfaces Monitoring of Mechanical Components manages change in material properties (Reduced thermal insulation resistance /

moisture intrusion) for Insulation component types comprised of Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber subjected to Air.Revise the first paragraph in the description section in LRA Sections A.1.18 and B.2.18, External Surfaces Monitoring of Mechanical Components, to address the aging effects applicable to this component, material, and environments as follows. A clarification is provided by separating structural commodities and mechanical components and the associated materials and aging effects:

From:

A.1.18 EXTERNAL SURFACES MONITORING OF MECHANICALCOMPONENTS PROGRAM The External Surfaces Monitoring of Mechanical Components Program is a new condition monitoring program that will manage aging effects of components fabricated from metallic and elastomeric materials through periodic visual inspection of external surfaces during system inspections and walkdowns for evidence of leakage, loss of material (including loss of material due to wear), and cracking, and change in material properties of elastomeric structural commodities. The visual inspection of elastomers will detect change in material properties such as cracking or crazing, swelling, discoloration and melting. Physical manipulation (of at least 10% of available surface area), such as touching, pressing, flexing, and bending, will be used to augment visual inspections to confirm the absence of hardening and loss of strength in elastomeric materials. The periodic inspections will include visual inspection of insulation jacketing to ensure the integrity of the jacketing is maintained.

To:

A.1.18 EXTERNAL SURFACES MONITORING OF MECHANICALCOMPONENTS PROGRAM The External Surfaces Monitoring of Mechanical Components Program is a new condition monitoring program that will manage aging effects of components fabricated from metallic, elastomeric, and insulating materials through periodic visual inspection of external surfaces during system inspections and walkdowns for evidence of leakage, loss of material (including loss of material due to wear) hardening and loss of strength

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 8 of 9 due to elastomer degradation, and cracking, in mechanical components. Furthermore, in structural commodities these inspections include change in material properties and cracking of elastomer due to elastomer degradation and change in material properties (causing reduction in thermal resistance) in insulating (i.e., Fiberglass/ Alumina silicate/

Calcium silicate/ Mineral fiber) materials and cracking and loss of material in aluminum structural commodities. The visual inspection of elastomers will detect change in material properties including cracking or crazing, swelling, discoloration and melting.

Physical manipulation (of at least 10% of available surface area), such as touching, pressing, flexing, and bending, will be used to augment visual inspections to confirm the absence of hardening and loss of strength in elastomeric materials. The periodic inspections will include visual inspection of insulation jacketing to ensure the integrity of the jacketing is maintained.

From:

B.2.18 EXTERNAL SURFACES MONITORING OF MECHANICAL COMPONENTS PROGRAM Program Description The External Surfaces Monitoring of Mechanical Components Program is a new condition monitoring program that will manage aging effects of components fabricated from metallic and elastomeric materials through periodic visual inspection of external surfaces during system inspections and walkdowns for evidence of leakage, loss of material (including loss of material due to wear), and cracking, and change in material properties of elastomer structural commodities. The visual inspection of elastomers will detect change in material properties including cracking or crazing, swelling, discoloration and melting. Physical manipulation (of at least 10% of available surface area), such as touching, pressing, flexing, and bending, will be used to augment visual inspections to confirm the absence of hardening and loss of strength in elastomeric materials. The periodic inspections will include visual inspection of insulation jacketing to ensure the integrity of the jacketing is maintained.

To:

B.2.18 EXTERNAL SURFACES MONITORING OF MECHANICAL COMPONENTS PROGRAM Program Description The External Surfaces Monitoring of Mechanical Components Program is a new condition monitoring program that will manage aging effects of components fabricated from metallic, elastomeric, and insulating materials through periodic visual inspection of external surfaces during system inspections and walkdowns for evidence of leakage, loss of material (including loss of material due to wear) hardening and loss of strength due to elastomer degradation, and cracking, in mechanical components. Furthermore, in structural commodities these inspections include change in material properties and cracking of elastomer due to elastomer degradation and change in material properties (causing reduction in thermal resistance) in insulating (i.e., Fiberglass/ Alumina silicate/

Calcium silicate/ Mineral fiber) materials and cracking and loss of material in aluminum structural commodities. The visual inspection of elastomers will detect change in

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 15 Page 9 of 9 material properties including cracking or crazing, swelling, discoloration and melting.

Physical manipulation (of at least 10% of available surface area), such as touching, pressing, flexing, and bending, will be used to augment visual inspections to confirm the absence of hardening and loss of strength in elastomeric materials. The periodic inspections will include visual inspection of insulation jacketing to ensure the integrity of the jacketing is maintained.

Intended Function - Fire Barrier See Question 3, Request 2 for components with a Fire barrier function.

5. Please discuss whether any of the non-fire barrier fiberglass/alumina silicate/calcium silicate/mineral fiber component types installed at Perry are located in areas of high gamma radiation. If they are, then discuss whether change in material properties is an applicable aging effect.

Fiberglass/alumina silicate/calcium silicate/mineral fiber Component type Penetration sealant (flood, radiation) is in areas of gamma radiation dose exceeding 10^6 rads. Change in material properties is an applicable aging effect.

PNPP will revise the following row in LRA Table 3.5.2-4 which previously has None for an aging effect:

Table 3.5.2-4 Component Type: Penetration sealant (flood, radiation)

Intended Function: EN, FLB, SPB, SNS, SRE, SHD Material: Fiberglass/ Alumina silicate/ Calcium silicate/ Mineral fiber Environment: Air - indoor, uncontrolled (Ext)

AERM: Change in material properties AMP: External Surfaces Monitoring of Mechanical Components NUREG-1801 Item: VIII.I.S-403 (LR-ISG-2012-02)

Table 1 Item: 3.4.1-64 (LR-ISG-2012-02)

Note: C Change LRA Note 503 to state Not Used.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 16 Page 1 of 4 6 NCSG RAI-10337-R1 Question 4 (Unimpregnated Fiberglass Fabric; Fiberglass Fabric Impregnated With Elastomer)

Background

LRA Tables 3.5.2-1 and 3.5.2-4 cite change in material properties and cracking for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer drywell mechanical penetration (fiberglass fabric), penetration sealant (fire), and safety relief valve (SRV) tailpipe penetration boot seals. LRA Supplement 3 (ML24206A150) revised LRA Tables 3.5.2-1 and 3.5.2-4 to also cite loss of material for these components.

LRA Supplement 3 (ML24206A150) revised LRA Table 3.5.2-4 to remove the Structures Monitoring program as one of the credited programs for managing change in material properties and cracking of unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer SRV tailpipe penetration boot seals.

Issue It is unclear why separation and delamination were not cited as applicable aging effects for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer drywell mechanical penetration (fiberglass fabric), penetration sealant (fire), and SRV tailpipe penetration boot seals.

Revision 6 of LRPY-CAMR-001 appears to only evaluate change in material properties and cracking and references Table 7-5 of EPRI 3002013084. The staff notes that Section 7 of EPRI 3002013084 states that [t]his section covers elastomers used in structural applications within the scope of LR, except for those associated with fire stops and penetration seals, which are addressed in Section 6. Therefore, based on Table 6-2 of EPRI 3002013084, separation due to vibration, movement, and shrinkage may be applicable to fire stops, and delamination may be applicable to elastomer fire stops.

Request

1. Please discuss why separation was not cited as an applicable aging effect for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer drywell mechanical penetration (fiberglass fabric), penetration sealant (fire), and SRV tailpipe penetration boot seals (e.g., does not interact with vibrating components, is not affected by differential movement between structures, etc.).

2. Please discuss why delamination was not cited as an applicable aging effect for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer drywell mechanical penetration (fiberglass fabric), penetration sealant (fire), and SRV tailpipe penetration boot seals.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 16 Page 2 of 4 PNPP Response

1. Please discuss why separation was not cited as an applicable aging effect for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer drywell mechanical penetration (fiberglass fabric), penetration sealant (fire), and SRV tailpipe penetration boot seals (e.g., does not interact with vibrating components, is not affected by differential movement between structures, etc.).

2. Please discuss why delamination was not cited as an applicable aging effect for unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer drywell mechanical penetration (fiberglass fabric), penetration sealant (fire), and SRV tailpipe penetration boot seals.

PNPP acknowledges the discrepancies associated with the component types of drywell mechanical penetration (fiberglass fabric), penetration sealant (fire), and SRV tailpipe penetration boot seals for the material unimpregnated fiberglass fabric; fiberglass fabric impregnated with elastomer. To correct the noted discrepancies, following actions will be taken:

The following new Table 2 rows will be added to LRA Table 3.5.2-1 and Table 3.5.2-4 as noted below:

Table 3.5.2-1 Component Type: Drywell mechanical penetration (fiberglass fabric)

Intended Function: EN,SPB,SSR, FB, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-1 Component Type: Drywell mechanical penetration (fiberglass fabric)

Intended Function: EN,SPB,SSR, FB, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking/Delamination AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 16 Page 3 of 4 Table 3.5.2-4 Component Type: Penetration sealant (fire)

Intended Function: EN, FB, FLB, SPB, SNS, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-4 Component Type: Penetration sealant (fire)

Intended Function: EN, FB, FLB, SPB, SNS, SRE Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking/Delamination AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 Table 3.5.2-4 Component Type: SRV Tailpipe Penetration Boot Seals Intended Function: EN, FB, FLB, SPB, SNS Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Separation AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 16 Page 4 of 4 Table 3.5.2-4 Component Type: SRV Tailpipe Penetration Boot Seals Intended Function: EN, FB, FLB, SPB, SNS Material: Unimpregnated fiberglass fabric; Fiberglass fabric impregnated with elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Cracking/Delamination AMP: Fire Protection NUREG-1801 Item: N/A Table 1 Item: N/A Note: H, 505 The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 17 Page 1 of 5 7 NCSG RAI-10337-R1 Question 5 (Gypsum Board/Drywall)

Background

LRA Supplement 3 (ML24206A150) revised LRA Table 3.5.2-4 to identify gypsum board drywall as a fire barrier used at Perry. However, no aging effects were cited, and plant-specific note 532 was added, which states, [n]o mechanism for degradation of drywall was identified due to aging.

Issue Based on Table 6-2 of EPRI 3002013084, loss of material due to abrasion may be an applicable aging effect for fire stops; cracking due to vibration, movement, and shrinkage may be an applicable aging effect for fire stops; change in materials due to gamma irradiation exposure may be an applicable aging effect for fire stops; and separation due to vibration, movement, and shrinkage may be an applicable aging effect for fire stops.

Revision 6 of LRPY-CAMR-001 appears to have only evaluated loss of material and change in material properties for gypsum board and did not address cracking or separation. LRPY-CAMR-001 states that loss of material due to abrasion is not applicable because it is due to design problems or human interaction and change in material properties due to gamma irradiation exposure is not applicable because it is not an organic polymer.

The staff notes that if gypsum board is experiencing loss of material due to design problems or human interaction, and corrective actions have not been taken to correct the design problems and human interaction, then loss of material is a known degradation that could impact the gypsum board from performing its intended fire barrier function if not managed.

Section 6.3.3.2 of EPRI 3002013084 states that 10^6 rads, which is the organic polymer radiation damage threshold, is conservative for fire wraps and fire stops and if the radiation dose is below 10^6 rads, then change in material properties is not expected. Table 6-3 of EPRI 3002013084 includes change in material properties due to gamma irradiation exposure for fire stops as being applicable for exposures exceeding 10^6 rads. Therefore, no explanation is provided for why change in material properties due to gamma radiation exposure is not applicable.

Request Please discuss why loss of material (e.g., not in vicinities of other vibrating components, design problems have been corrected, and human interaction problems have been corrected), change in material properties due to gamma irradiation (e.g., not located in areas where the gamma irradiation exposure would exceed 10^6 rads), cracking (e.g., not in vicinities of other vibrating components), and separation (e.g., not in vicinities of other vibrating components) are not applicable aging effects for gypsum board.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 17 Page 2 of 5 PNPP Response Please discuss why loss of material (e.g., not in vicinities of other vibrating components, design problems have been corrected, and human interaction problems have been corrected), change in material properties due to gamma irradiation (e.g., not located in areas where the gamma irradiation exposure would exceed 10^6 rads), cracking (e.g., not in vicinities of other vibrating components), and separation (e.g., not in vicinities of other vibrating components) are not applicable aging effects for gypsum board.

PNPP Response is organized by providing general information including that discussed in the breakout session on TRP-026-06 and then divides this question in four parts addressing each aging effect separately.

LRA Table 3.5.2-4 identifies the component type drywall made from gypsum board and having a fire barrier intended function. Drywall is used as a rigid means to divide areas not exposed to weather (e.g. internal walls) supporting the intended function fire barrier. These fire barriers offer reasonable assurance that a fire will not spread from one plant area to another. EPRI Report 3002013084 Section 5 discusses concrete walls but not walls made out of gypsum. Section 6 evaluates aging effects associated with fire wraps and fire stops. To evaluate these aging effects PNPP considers these gypsum board barriers as a large, rigid fire stops for the purpose of discussing aging effects. Typically fire stops are small and fill gaps in larger commodities but is used here to identify mechanisms that could cause aging effects in gypsum board/drywall.

EPRI Report 3002013084 sections 6.3.1 and 6.3.2 do not classify damage of gypsum board drywall due to design problems having been corrected, and human interaction problems as aging mechanisms causing the aging effects of Loss of material and cracking. However, it does classify damage due to being in the vicinities of other vibrating components as an aging mechanism causing the aging effects of loss of material and cracking / delamination. The same information is contained in the EPRI Structural Tools report for initial license renewal No.

1015078. These conditions are event driven mechanisms for loss of material and cracking and are not the subject of license renewal. The current license basis requires there be a Fire Protection Program that inspects gypsum board assemblies that are used as fire barriers, and these inspections address the damage from event-based mechanisms, as well as damage due to vibration, and make corrections when the damage exceeds acceptance criteria. These inspections will continue through the PEO, as inherently required by Condition 2(c)6 of the facility operating license. The Fire Protection Program establishes a robust set of criteria that would identify the degradation of fire-rated gypsum board assemblies resulting from non-age-related mechanisms and is enhanced, if necessary, to incorporate detection of aging effects.

The applicable procedures were provided to the NRC during the break-out sessions audit and formed the bases for addressing gypsum board in a comprehensive way.

A keyword search for the material gypsum or drywall did not result in any occurrences when searching any of: NUREG 1801 R2, NUREG 2191 Volumes 1 and 2, and LR-ISG-2021-02 or LR-ISG 2021-03. The basis for this discussion on gypsum board adopts the information in EPRI structural tools for the initial License Renewal and subsequent license renewal plants.

PNPPs bases document addresses Loss of material and Change in material properties but did not explicitly disposition cracking and separation. The basis document also did not

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 17 Page 3 of 5 explicitly disposition the aging mechanism of vibration (considered the aging mechanism causing cracking in the EPRI structural tools), and in addition, the cause of abrasion.

In all cases, these aging mechanisms are not applicable. PNPP acknowledges that the basis document does not explicitly disposition these aging mechanisms/effects; however, this has no impact on the PNPP LRA content.

Please discuss why loss of material (e.g., not in vicinities of other vibrating components, design problems have been corrected, and human interaction problems have been corrected) is not applicable aging effects for gypsum board.

In the structural tools, the aging mechanisms for loss of material are flaking, scouring and abrasion. Flaking and scouring were considered but neither is considered applicable to fire stops. Abrasion results when gypsum board/drywall encounters the effects of vibrating equipment. During the PEO there is reasonable assurance that this damage would not occur.

Vibration induced damage would require equipment to be in direct contact with rated assemblies, or (in the case of piping/ductwork) penetrating the assembly with no penetration seal. Such conditions would be atypical of general plant design and would represent design issues. If the vibration were present, it is reasonable based upon experience that it would have already been experienced and corrected during the initial 40 years of operation. As such, vibration issues causing abrasion in fire stops are considered design issues or human interaction problems, e.g. from maintenance activities. Both these issues and problems are not aging mechanisms and as such not subject to aging management.

In summary, PNPP bases document and EPRI Structural tools conclude that:

Gypsum board drywall component type is a fire barrier treated as a fire stop for the purposes of identifying potential aging mechanisms.

The aging mechanisms that may cause loss of material for fire wrap and fire stop fire barriers are flaking, scouring, and abrasion.

Of these mechanisms abrasion from fire stops in the vicinity of vibrating equipment for gypsum board could cause loss of material.

PNPP considers this condition as being due to a design problem or human interaction and would have been discovered and corrected because of the current licensing basis Fire Protection Program requirements.

Ongoing human interaction problems are not age related mechanisms.

In the absence of these aging mechanisms, loss of material is not an applicable aging effect.

No LRA changes are necessary because of this part of the question.

Please discuss why change in material properties due to gamma irradiation (e.g., not located in areas where the gamma irradiation exposure would exceed 10^6 rads) is not applicable aging effects for gypsum board.

At PNPP, the component type Drywall made from gypsum board is only used as a fire barrier in mild environments. Mild environments do not contain radiation dose that is expected to exceed the threshold dose cited for the amount of irradiation necessary for change in material

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 17 Page 4 of 5 properties to be applicable throughout the PEO. Consequently, in the absence of any consequential radiation dose, change in material properties is not an applicable aging effect.

No LRA changes are required due to this part.

Please discuss why cracking (e.g., not in vicinities of other vibrating components) is not applicable aging effects for gypsum board.

Per the structural tools, the aging mechanisms that may cause cracking or delamination in fire wrap and fire stops are vibration, movement, and shrinkage. Cracking was not evaluated for gypsum board fire barriers in the PNPP bases document.

Per Section 6.3.2.2 of the Structural Tools, differential movement of fire stops between structures may cause cracking. Being a bulk commodity per LRA Table 3.5.2-4, gypsum board/drywall fire barriers are not used for separating structures. It is only used as internal fire barriers within structures. Therefore, movement is not an applicable aging mechanism at PNPP.

Per Section 6.3.2.3, Structural Tools, Shrinkage may occur over time where fire penetration seal material comes in contact with pipe surfaces. Gypsum board/drywall is not a fire penetration seal material and therefore shrinkage is not applicable.

At PNPP, the component type Drywall made from gypsum board is only used as a rigid fire barrier for interior walls. The basis for determining the aging effect of cracking is from the information in EPRI structural tools Report 3002013084, Section 6.3.1. The source identifies aging mechanisms of vibration, movement, and shrinkage. As discussed in the general information vibration issues causing cracking would have been due to a design program and eliminated early in the plants life. EPRI considers movement between adjacent structures.

However, drywall is used as a bulk commodity and is an internal barrier not separating structures. Hence, movement is not an applicable mechanism.

Structural Tools, Section 6.3.2.1, lists vibration as a potential aging mechanism causing cracking or delamination of fire wrap and fire stop material over time. It goes on to explain for flexible fire stop materials in the vicinity of vibrating equipment inducing vibrations in the material may cause delamination or fatigue based splits. As discussed in the section on loss of material above, PNPP considers vibration induced environment for gypsum board/drywall a design problem or it could be caused by human interaction. Both mechanisms would have been discovered and corrected in the current licensing basis fire protection program, eliminating such issues for the PEO. Human interaction causes, like design issues, are considered event driven and not an aging mechanism.

In summary, PNPP bases document and EPRI Structural tools conclude that:

Gypsum board/drywall component type is a fire barrier treated as a fire stop for the purposes of identifying potential aging mechanisms.

The aging mechanisms that may cause cracking for fire wrap and fire stop fire barriers are vibration, movement, and shrinkage.

Of these mechanisms vibration in the vicinity of vibrating equipment for gypsum could cause cracking, whereas movement and shrinkage is not applicable to gypsum board/drywall fire stops.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 17 Page 5 of 5 PNPP considers vibration as being due to a design problem or human interaction and would have been discovered and corrected over the prior 40 years because of the current licensing basis Fire Protection program that is required to inspect fire barriers. Hence, during the PEO there is reasonable assurance that cracking due to vibration would not occur.

Ongoing human interaction problems are not age related mechanisms.

In the absence of vibration, movement, and shrinkage aging mechanisms, Cracking is not an applicable aging effect.

Please discuss why separation (e.g., not in vicinities of other vibrating components) is not applicable aging effects for gypsum board.

Per the structural tools, Separation is the destruction of adhesion between a fire barrier material and an adjacent surface. The aging mechanisms that may cause cracking or delamination in fire wrap and fire stops are vibration, movement, and shrinkage. Gypsum board is not installed as an adhesive. Hence, Separation due to vibration, movement, and shrinkage for gypsum board/drywall is not an applicable aging effect.

In summary, in the absence of any of aging mechanisms namely vibration, movement, and shrinkage, the associated aging effect of Separation is not applicable.

No LRA changes are necessary as a result of this part of the question.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 18 Page 1 of 2 8 NCSG RAI-10337-R1 Question 6 (Loss of Sealing)

Background

LRA Supplement 3 (ML24206A150) revised LRA Table 3.5.2-4 to delete loss of sealing of elastomer fire stops managed by the Structures Monitoring program. Only the Fire Protection program is credited for managing loss of sealing of the elastomer fire stops. In addition, LRA Table 3.5.2-4 credits only the Fire Protection program for managing loss of sealing of the elastomer seismic isolation joints. Plant-specific note 522 is cited and states, Structures Monitoring Program is aligned with Fire Protection Program in detecting the loss of sealing aging effect for this material/environment combination. This plant-specific note appears to indicate that the Structures Monitoring and Fire Protection programs work together to manage loss of sealing of the elastomer fire stops and seismic isolation joints.

The staff notes that LRA Table 3.5.2-4 does credit both the Fire Protection and Structures Monitoring programs for managing loss of sealing of elastomer penetration sealant (fire), also citing plant-specific note 522. This approach is consistent with the approach for elastomer shield building electrical penetration seals and sealant in LRA Table 3.5.2-1, which credits both the 10 CFR 50, Appendix J, and Fire Protection programs with managing loss of sealing. The approach of crediting two programs seems appropriate given the components have intended functions, in addition to the fire barrier intended function.

Issue It is unclear why, consistent with what appears to be the intent of plant-specific note 522, both the Fire Protection and Structures Monitoring programs are credited to manage loss of sealing of the elastomer penetration sealant (fire), but both programs are not credited to manage loss of sealing of the elastomer fire stops and seismic isolation joints, which also cite plant-specific note 522.

Request Please discuss why both the Fire Protection and Structures Monitoring programs are credited to manage loss of sealing of the elastomer penetration sealant (fire), but both programs are not credited to manage loss of sealing of the elastomer fire stops and seismic isolation joints.

PNPP Response Please discuss why both the Fire Protection and Structures Monitoring programs are credited to manage loss of sealing of the elastomer penetration sealant (fire), but both programs are not credited to manage loss of sealing of the elastomer fire stops and seismic isolation joints.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 18 Page 2 of 2 PNPP acknowledges the discrepancy in identifying the AMPs used for the component types of Penetration sealants (fire), fire stops and seismic isolation joint for the material elastomer associated with aging effect loss of sealing. In order to correct these discrepancies, the following actions will be taken:

Plant-specific Note 522 will be revised to state as follows:

Fire Protection Program will detect the aging effect loss of sealing for this material/environment combination.

The following row from LRA Table 3.5.2-4 will be removed:

Table 3.5.2-4 Component Type: Penetration sealant (fire)

Intended Function: EN, FB, FLB, SPB, SNS, SRE Material: Elastomer Environment: Air - indoor, uncontrolled (Ext)

AERM: Loss of sealing AMP: Structures Monitoring NUREG-1801 Item: III.A6.TP-7 Table 1 Item: 3.5.1-72 Note: A The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 19 Page 1 of 4 9 NCSG RAI-10339-R1 Regulatory Basis 10 CFR 54.21(a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

Question 1 - RAI B.2.45-1 Introduction As added by letter dated September 5, 2024 (ML24249A123), LRA Section B.2.45 describes the new Plant-Specific Periodic Inspections for Selective Leaching Program as plant-specific. The applicant and staff identified three populations (i.e., materials and environment combinations) where selective leaching is occurring, and the applicant provided the Plant-Specific Periodic Inspections for Selective Leaching Program to manage loss of material due to selective leaching for these populations. The three populations being managed using this plant-specific AMP are:

(a) gray cast iron components exposed to raw water; (b) gray cast iron components exposed to soil; and (c) ductile iron components exposed to soil. With the issuance of GALL-SLR Report AMP XI.M33, Selective Leaching, the staff provided a framework to manage loss of material due to selective leaching through periodic inspections, as opposed to the GALL-LR Report AMP XI.M33 framework which recommends one-time inspections to demonstrate that this aging effect is not occurring. In addition, the staff noted the applicant developed the Plant-Specific Periodic Inspections for Selective Leaching Program based on the guidance provided in GALL-SLR Report AMP XI.M33. Therefore, the staff compared the program elements included in LRA Section B.2.45 to the corresponding program elements of GALL-SLR Report AMP XI.M33.

Background

The detection of aging effects program element in LRA Section B.2.45 states the following (in part):

[m]echanical examination techniques, such as chipping and scraping, augmented visual inspections for gray cast iron components.

[t]he Fire Protection System contains one gray cast iron (ductile iron) piping component with a soil environment. For populations with less than 35 components, PNPP will perform one destructive examination for this population during each inspection period The acceptance criteria program element in LRA Section B.2.45 states the following (in part):

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 19 Page 2 of 4 b) the presence of no more than a superficial layer of dealloying, as determined by removal of the dealloyed material by mechanical removal, and c) the components meet system design requirements such as minimum wall thickness, when projected to the end of the period of extended operation. When evaluating a component in relation to criterion (b) [emphasis added by staff], no credit is taken for the material properties of the dealloyed portion of the component.

GALL-SLR Report AMP XI.M33 recommends the following:

Mechanical examination techniques, such as chipping and scraping, augment visual inspections for gray cast iron and ductile iron [emphasis added by staff] components.

Two destructive examinations for sample populations with greater than 35 susceptible components. When inspections are conducted on piping, a 1-foot axial length section is considered as one inspection.

Issue

1. The staff seeks clarification with respect to why mechanical examination techniques, such as chipping and scraping, augment visual inspections for gray cast iron components but not ductile iron components.

2. It is the staffs understanding that there is greater than 35 feet of in-scope ductile iron piping exposed to soil. Based on this, the staff seeks clarification with respect to why two destructive examinations will not be performed for this population.

3. The staff seeks clarification with respect to why criterion (b) (highlighted and italicized by the staff in the background section above) does not refer to criterion (c), Criterion (b) refers to a superficial dealloyed layer which would not involve an evaluation, whereas criterion (c) would involve an evaluation to show that system design requirements would be met.

Request

1. State the basis for why mechanical examination techniques will not augment visual inspections for ductile iron components. Alternatively, revise the LRA to reflect that mechanical examination techniques will augment visual inspections for ductile iron components.

2. Clarify if there is greater than 35 feet of in-scope ductile iron piping exposed to soil. If there is, state the basis for performing only one destructive examination for this population during each inspection period. Alternatively, revise the LRA to reflect that two destructive examinations will be performed for this population during each inspection period.

3. State the basis for why criterion (b) (highlighted and italicized by the staff in the background section above) does not refer to criterion (c). Alternatively, revise the LRA to reference criterion (c) instead of criterion (b).

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 19 Page 3 of 4 PNPP Response

1. State the basis for why mechanical examination techniques will not augment visual inspections for ductile iron components. Alternatively, revise the LRA to reflect that mechanical examination techniques will augment visual inspections for ductile iron components.

PNPP agrees that clarifications to the enhancements for Element 4, Detection of Aging Effects, of LRA Section B.2.45 are appropriate.

Bullet 2 of the first set of bullets in Element 4 will be revised as follows:

From:

Mechanical examination techniques, such as chipping and scraping, augmented visual inspections for gray cast iron components.

To:

Mechanical examination techniques, such as chipping and scraping, augmented visual inspections for gray cast iron and ductile iron components.

2. Clarify if there is greater than 35 feet of in-scope ductile iron piping exposed to soil. If there is, state the basis for performing only one destructive examination for this population during each inspection period. Alternatively, revise the LRA to reflect that two destructive examinations PNPP agrees that clarification to the enhancements for Element 4, Detection of Aging Effects, of LRA Section B.2.45 is appropriate.

The bulleted text regarding destructive examinations in Element 4 will be revised as follows:

From:

The Fire Protection System contains one gray cast iron (ductile iron) piping component with a soil environment. For populations with less than 35 components, PNPP will perform one destructive examination for this population during each inspection period; otherwise, a technical justification of the methodology and sample size used for selecting components for inspection will be included as part of the programs documentation.

To:

The Fire Protection System contains one gray cast iron (ductile iron) piping component type with a soil environment containing more than 35 linear feet of in-scope pipe; 1-foot axial lengths are considered in the population determination. For populations with more than 35 1-foot axial lengths, PNPP will perform two destructive examinations for this population during each inspection period; otherwise, a technical justification of the methodology and sample size used for selecting components for inspection will be included as part of the programs documentation.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 19 Page 4 of 4

3. State the basis for why criterion (b) (highlighted and italicized by the staff in the background section above) does not refer to criterion (c). Alternatively, revise the LRA to reference criterion (c) instead of criterion (b).

NUREG-2191 XI.M33 Acceptance Criteria contains items a) through d). Item a) applies only to copper pipe. PNPP does not have buried copper pipe so item a) was not included, the remaining three items were renumbered a) through c) such that item c) in NUREG2191 XI.M33 Section 2.6 was presented as item b). PNPP acknowledges that this likely introduces unnecessary confusion. To promote consistency with NUREG-2191, the Element 6 text will be revised as follows (applicable portion only presented here):

Element 6 - Acceptance Criteria: The acceptance criteria for gray cast iron and ductile iron will include:

a) Criterion (a) of NUREG-2191, XI.M33 is not applicable to PNPP as PNPP does not have buried copper pipe, b) for gray cast iron and ductile iron, the absence of a surface layer that can be easily removed by chipping or scraping or identified in the destructive examinations, c) the presence of no more than a superficial layer of dealloying, as determined by removal of the dealloyed material by mechanical removal, and d) the components meet system design requirements such as minimum wall thickness, when projected to the end of the period of extended operation.

When evaluating a component in relation to criterion (c) no credit is taken for the material properties of the dealloyed portion of the component.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 20 Page 1 of 2 0 NCSG RAI-10339-R1 Question 2 - RAI A.1.45-1 Introduction As added by letter dated September 5, 2024, LRA Section A.1.45 provides the UFSAR supplement for the Plant-Specific Periodic Inspections for Selective Leaching Program. The applicant and staff identified three populations (i.e., materials and environment combinations) where selective leaching is occurring, and the applicant provided the Plant-Specific Periodic Inspections for Selective Leaching Program to manage loss of material due to selective leaching for these populations. The three populations being managed using this plant-specific AMP are:

(a) gray cast iron components exposed to raw water; (b) gray cast iron components exposed to soil; and (c) ductile iron components exposed to soil. With the issuance of GALL-SLR Report AMP XI.M33, Selective Leaching, the staff provided a framework to manage loss of material due to selective leaching through periodic inspections, as opposed to the GALL-LR Report AMP XI.M33 framework which recommends one-time inspections to demonstrate that this aging effect is not occurring. In addition, the staff noted the applicant developed the Plant-Specific Periodic Inspections for Selective Leaching Program based on the guidance provided in GALL-SLR Report AMP XI.M33. Therefore, the staff compared LRA Section A.1.45 against the recommended description for AMP XI.M33 as described in GALL-SLR Report Table XI-01, FSAR Supplement Summaries for GALL-SLR Report Chapter XI Aging Management Programs.

Background

The recommended description for AMP XI.M33 as described in GALL-SLR Report Table XI-01 states [w]hen the acceptance criteria are not met such that it is determined that the affected component should be replaced prior to the end of the (Delete: subsequent) [deleted by staff since this is an initial LRA] period of extended operation, additional inspections are performed.

Issue The staff seeks clarification with respect to why the statement from GALL-SLR Report Table XI-01 (described in the background section above) is not included in LRA Section A.1.45.

Request State the basis for why the statement from GALL-SLR Report Table XI-01 (described in the background section above) is not included in LRA Section A.1.45. Alternatively, revise LRA Section A.1.45 to include this statement.

PNPP Response State the basis for why the statement from GALL-SLR Report Table XI-01 (described in the background section above) is not included in LRA Section A.1.45. Alternatively, revise LRA Section A.1.45 to include this statement.

Perry Nuclear Power Plant Responses to LRA NRC RAIs Set 3 L-24-209 Attachment 20 Page 2 of 2 The intent of the statement quoted in the Background section above is addressed in the content of Appendix B of the PNPP LRA, as modified by Supplement 4 of the LRA (Vistra Letter L 200), specifically the requirements provided for Elements 5 and 6 of the new plant-specific program. However, PNPP acknowledges that the additional guidance from GALL-SLR, incorporated into LRA Appendix A (and therefore the future UFSAR supplement) provides additional information that would improve the Plant-Specific Periodic Inspections for Selective Leaching Program. Consequently, LRA Section A.1.45, third paragraph, will be revised to read:

The Plant-Specific Periodic Inspections for Selective Leaching Program includes periodic and opportunistic inspections using visual examinations coupled with mechanical examination techniques, and destructive examinations. These techniques can determine whether loss of materials due to selective leaching is occurring and whether selective leaching will affect the ability of the components to perform their intended function for the period of extended operation. When the acceptance criteria are not met such that it is determined that the affected component should be replaced prior to the end of the period of extended operation, additional inspections are performed.

The proposed updates to PNPP LRA addressed in this RAI response will be submitted as part of a later supplement to the LRA.

References None Attachments None

L-24-209, License Renewal Application for the Perry Nuclear Power Plant - Responses to Request for Additional Information - Round 1 (Set 3)

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L-24-209, License Renewal Application for the Perry Nuclear Power Plant - Responses to Request for Additional Information - Round 1 (Set 3)

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