L-24-179, License Renewal Application for the Perry Nuclear Power Plant Revision 0 - Supplement 5
| ML24295A352 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 10/21/2024 |
| From: | Penfield R Vistra Corp |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-24-179 | |
| Download: ML24295A352 (1) | |
Text
v1c-r1:.1 11111111117f ll!lila L-24-179 October 21, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry Nuclear Power Plant, Unit No. 1 Docket No. 50-440, License No. NPF-58 Perry Nuclear Power Plant Rod L. Penfield Site Vice President 10 Center Road Perry, Ohio 44081 10 CFR 54 License Renewal Application for the Perry Nuclear Power Plant Revision O - Supplement 5
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. ML23184A081)
- 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)
Perry Nuclear Power Plant L-24-179 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 O, 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. NRC Email from Vaughn Thomas to Rod Penfield - dated August 14, 2024 - Perry LRA -
Requests for Additional Information - Set 1 (ADAMS Accession No. ML24227A956 and ML24227A957)
- 10. 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)
- 11. NRC Email from Vaughn Thomas to Rod Penfield - dated August 28, 2024 - Perry LRA -
Requests for Additional Information -Set 2 (ADAMS Accession No. ML24241A100 and ML24241A101)
- 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)
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 8).
In addition, as a result of the NRC's review and audit of the PNPP LRA, on August 14, 2024, the NRC Staff has submitted to the PNPP Staff the first and second sets of several requests for additional information (RAls) (References 9 and 11), which the PNPP Staff responded to on September 16, 2024, and October 2, 2024 via Vistra Letters L-24-207 (RAI Set 1 Responses - Reference 10) and Vistra Letter L-24-208 (RAI Set 2 Responses - Reference 12).
Attachments 1 to 7 of this letter is Supplement 5 of the PNPP LRA, which provides miscellaneous LRA updates to address previous NRC Staff issues from the LRA audit or RAls.
0
Perry Nuclear Power Plant L-24-179 Page 3 of 3 For ease of reference, an index listing the associated attachments is provided.
The commitments provided in the PNPP LRA Appendix A (Table A.3) that are supplemented are indicated in the attachments. 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 October 21, 2024.
Attachments:
NRC Region Ill Administrator NRC Resident Inspector NRR Project Manager Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)
Utility Radiological Safety Board
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachments Index Page 1 of 2 PNPP LRA Supplement 5 Attachments Index for Miscellaneous Supplemented LRA Sections and Tables Attachment No.
LRA Section, Table or Appendix Supplemented Subject Source 1
Table 3.1.2-2 Bolting Integrity aging management program update Applicant Initiated 2
Section 4.5.3 Update for Fatigue Monitoring aging management program to add containment piping penetrations bellows TRP-063-03 3
Section A.1.37 Added enhancement for Open-Cyle Cooling Water System aging management program for heat exchanger thermal performance monitoring NCSG RAI-10183-R1 4
Section A.2.5.3 Update for Fatigue Monitoring aging management program to add containment piping penetrations bellows TRP-063-03 5
Appendix A Table A.3 Added enhancement for Open-Cyle Cooling Water System aging management program for heat exchanger thermal performance monitoring TRP-063-03 NCSG RAI-10183-R1 6
Appendix B Table B.1-2 Added enhancement for Open-Cyle Cooling Water System aging management program for heat exchanger thermal performance monitoring Applicant Initiated NCSG RAI-10183-R1 7
Appendix B Section B.2.37 Added enhancement for Open-Cyle Cooling Water System aging management program for heat exchanger thermal performance monitoring and editorial change NCSG RAI-10183-R1
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachments Index Page 2 of 2 The attachments incorporate the Perry Nuclear Power Plant LRA changes made via the LRA supplements, the annual update and RAI responses which were submitted via the following Vistra correspondence:
Therefore, the LRA updates depicted in the attachments are made on clean LRA pages that reflect the LRA updates from the previously docketed Vistra correspondence listed above.
Revisions to LRA tables may be shown by providing excerpts from each affected table, i.e., only the affected parts of the table may be included in the attachment.
Consistent with LRA supplements and the annual update, changes for the attachments are indicated by red, bolded and underlined text for added text and strikethrough for text to be deleted.
Note that text editing changes to some of the attachments such as spacing, font consistency changes etc., are not indicated via coloring as these are inconsequential.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 1 of 10 LRA Section: Table 3.1.2-2 LRA Page Number(s): 3.1-47 through 3.1-55
Reference:
Vistra Letter L-24-020 (PNPP LRA Supplement 2)
Description of Change: This LRA change is self-identified and voluntary (applicant initiated).
PNPP Bolting Integrity aging management program contains exceptions to the program in NUREG-1801, Revision 2. While updating Standard notes from A or C to B and D, respectively, in Supplement 2 of the PNPP LRA, the standard note in LRA Table 3.1.2-2 Row 5 was inadvertently changed, and the standard note in Row 7 was not changed when it should have been.
In Subsequent Responses, Rows 51 and 53 in LRA Table 3.1.2-2 were inserted incorrectly, and as such, given the wrong row numbers. This Attachment sorts these rows correctly.
Additionally, in Row 51, the environment Air - indoor, uncontrolled was missing (Ext), which signifies it as the environment associated with the external surfaces of the Valve bodies. This omission is corrected.
PNPP LRA Table 3.1.2-2, as previously modified by PNPP LRA Supplement 2 (Vistra Letter L-24-020) Pages 3.1-47 through 3.1-55 is revised as follows:
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 2 of 10 Table 3.1.2-2 Reactor Vessel, Internals and Reactor Coolant Systems - Nuclear Boiler Summary of Aging Management Evaluation Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 1
Accumulator Pressure boundary Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 2
Accumulator Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
3 Bolting Leakage boundary Steel Air - indoor, uncontrolled (Ext)
Loss of material Bolting Integrity IV.C1.RP-42 3.1.1-63 B
4 Bolting Leakage boundary Steel Air - indoor, uncontrolled (Ext)
Loss of preload Bolting Integrity IV.C1.RP-43 3.1.1-67 B
5 Bolting Pressure boundary Steel Air - indoor, uncontrolled (Ext)
Cumulative fatigue damage TLAA IV.C1.RP-44 3.1.1-11 B A 6
Bolting Pressure boundary Steel Air - indoor, uncontrolled (Ext)
Loss of material Bolting Integrity IV.C1.RP-42 3.1.1-63 B
7 Bolting Pressure boundary Steel Air - indoor, uncontrolled (Ext)
Loss of preload Bolting Integrity IV.C1.RP-43 3.1.1-67 A B 8
Filter housing Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 3 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 9
Filter housing Leakage boundary Stainless steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
10 Flexible hose Pressure boundary Nickel alloy Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 111 11 Flexible hose Pressure boundary Nickel alloy Air - indoor, uncontrolled (Ext)
None None IV.E.RP-03 3.1.1-106 A
12 Flexible hose Pressure boundary Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 13 Flexible hose Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
14 Flexible hose Pressure boundary Stainless steel Air - indoor, uncontrolled (Int)
None None VII.J.AP-123 3.3.1-120 A
15 Orifice Flow restriction, Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
16 Orifice Flow restriction, Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Cracking Water Chemistry and One-Time Inspection VIII.E.SP-88 3.4.1-11 B
17 Orifice Flow restriction, Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 4 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 18 Orifice Flow restriction, Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
19 Orifice Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
20 Orifice Leakage boundary Stainless steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
21 Piping Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
22 Piping Leakage boundary Stainless steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
23 Piping Leakage boundary Steel Air - indoor, uncontrolled (Ext)
Loss of material External Surfaces Monitoring of Mechanical Components V.E.E-44 3.2.1-40 A
24 Piping Leakage boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
25 Piping Leakage boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 5 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 26 Piping Leakage boundary Steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection V.D2.EP-60 3.2.1-16 B
27 Piping Pressure boundary Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 28 Piping Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
29 Piping Pressure boundary Stainless steel Air - indoor, uncontrolled (Int)
None None VII.J.AP-123 3.3.1-120 A
30 Piping Pressure boundary Stainless steel Treated water (Ext)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
31 Piping Pressure boundary Stainless steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
32 Piping Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Cracking Water Chemistry and One-Time Inspection VIII.E.SP-88 3.4.1-11 B
33 Piping Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 6 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 34 Piping Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
35 Piping Pressure boundary Steel Air - indoor, uncontrolled (Ext)
Loss of material External Surfaces Monitoring of Mechanical Components V.E.E-44 3.2.1-40 A
36 Piping Pressure boundary Steel Air - indoor, uncontrolled (Int)
Loss of material Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components V.D2.E-29 3.2.1-44 A
37 Piping Pressure boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
38 Piping Pressure boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
39 Piping Pressure boundary Steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection V.D2.EP-60 3.2.1-16 B
40 Piping Structural integrity Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 41 Piping Structural integrity Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 7 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 42 Piping Structural integrity Steel Air - indoor, uncontrolled (Ext)
Loss of material External Surfaces Monitoring of Mechanical Components V.E.E-44 3.2.1-40 A
43 Piping Structural integrity Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
44 Piping Structural integrity Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
45 Piping Structural integrity Steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection V.D2.EP-60 3.2.1-16 B
46 SRV Discharge Quencher Pressure boundary Stainless steel Treated water (Ext)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
47 SRV Discharge Quencher Pressure boundary Stainless steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
48 Strainer body Leakage boundary Steel Air - indoor, uncontrolled (Ext)
Loss of material External Surfaces Monitoring of Mechanical Components V.E.E-44 3.2.1-40 A
49 Strainer body Leakage boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 8 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 50 Strainer body Leakage boundary Steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection V.D2.EP-60 3.2.1-16 B
53 51 Valve Body Leakage Boundary Copper alloy >15%
ZN Air - indoor, uncontrolled (Ext)
None None V.F.EP-10 3.2.1-57 A
52 Valve Body Leakage Boundary Copper alloy >15%
ZN Treated water (Int)
Loss of material Selective Leaching VII.E3.AP-32 3.3.1-72 B
51 53 Valve Body Leakage Boundary Copper alloy >15%
ZN Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection VII.E3.AP-140 3.3.1-22 B
54 Valve body Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
55 Valve body Leakage boundary Stainless steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
56 Valve body Leakage boundary Steel Air - indoor, uncontrolled (Ext)
Loss of material External Surfaces Monitoring of Mechanical Components V.E.E-44 3.2.1-40 A
57 Valve body Leakage boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
58 Valve body Leakage boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 9 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 59 Valve body Leakage boundary Steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection V.D2.EP-60 3.2.1-16 B
60 Valve body Pressure boundary Aluminum Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-134 3.3.1-113 E, 105 61 Valve body Pressure boundary Aluminum Air - indoor, uncontrolled (Ext)
None None VII.J.AP-135 3.3.1-113 A
62 Valve body Pressure boundary Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 63 Valve body Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
64 Valve body Pressure boundary Stainless steel Air - indoor, uncontrolled (Int)
None None VII.J.AP-123 3.3.1-120 A
65 Valve body Pressure boundary Stainless steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
66 Valve body Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Cracking Water Chemistry and One-Time Inspection VIII.E.SP-88 3.4.1-11 B
67 Valve body Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 1 Page 10 of 10 Table 3.1.2 Nuclear Boiler System Row Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes 68 Valve body Pressure boundary Stainless steel Treated water
>60°C (>140°F)
(Int)
Loss of material Water Chemistry and One-Time Inspection IV.C1.RP-158 3.1.1-79 B
69 Valve body Pressure boundary Steel Air - indoor, uncontrolled (Ext)
Loss of material External Surfaces Monitoring of Mechanical Components V.E.E-44 3.2.1-40 A
70 Valve body Pressure boundary Steel Air - indoor, uncontrolled (Int)
Loss of material Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components V.D2.E-29 3.2.1-44 A
71 Valve body Pressure boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
72 Valve body Pressure boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
73 Valve body Pressure boundary Steel Treated water (Int)
Loss of material Water Chemistry and One-Time Inspection V.D2.EP-60 3.2.1-16 B
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 2 Page 1 of 2 LRA Section: Section 4.5.3 LRA Page Number(s): Page 4.5-3
Reference:
TRP-063-03, Vistra Letter L-24-189 (PNPP LRA Supplement 1)
Description of Change: In the response to the reference TRP in Vistra letters L-24-189 (Attachment 15) and L-24-020 (Attachments 42, 46, and 52), PNPP agreed to add the aging management of the containment piping penetrations bellows into the Fatigue Management Program. PNPP LRA Section 4.5-3 (as previously modified by LRA Supplement 1) is revised to reflect the results of the fatigue analyses performed to incorporate the containment piping penetrations bellows into the PNPP Fatigue Management Program.
PNPP LRA, Section 4.5.3 (as previously modified by PNPP Supplement 1 (Vistra Letter L 189), Page 4.5-3, is revised as follows:
4.5.3 CONTAINMENT PIPING PENETRATION BELLOWS TLAA
Description:
Guard pipe assemblies associated with containment penetrations utilize bellows. The PNPP specification required these bellows to be analyzed for at least 500 cycles of normal operation plus one safe shutdown earthquake (SSE) cycle for 40 years of operation.
Therefore, these fatigue analyses are identified as TLAAs requiring disposition for license renewal.
TLAA Evaluation:
PNPP has evaluated the containment piping penetrations bellows fatigue and determined, based on the 40-year CUF for the bounding penetration bellows, that the bellows fatigue usage is bounded by the fatigue usage of the penetrations. Therefore, since the penetration fatigue is not expected to exceed allowable limits at 60-years neither would the bellows fatigue.
However, since the PNPPs method for managing fatigue is to track and evaluate transient cycles and to calculate CUFs to ensure that the CUFs for the limiting components do not exceed design limits, therefore, fatigue analyses are performed for the containment piping penetrations bellows. These analyses are evaluated to select the highest fatigue usage bellows for monitoring. These high fatigue usage bellows bound or represent all other the containment piping penetrations bellows.
The Fatigue Monitoring Program calculates cumulative usage factors (CUFs) for the limiting locations and requires corrective actions if design limits are approached. This is accomplished by the use of cycle-based fatigue (CBF) monitoring where fatigue is computed from counted transients and parameters to ensure that the CUFs for the limiting locations do not exceed the design limit of 1.0.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 2 Page 2 of 2 Tthe effects of fatigue on the containment piping penetrations bellows will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21(c)(1)(iii).
Disposition:
The effects of fatigue on the containment piping penetrations bellows will be managed for the period of extended operation by the Fatigue Monitoring Program.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 3 Page 1 of 2 LRA Section: Appendix A, A.1.37 LRA Page Number(s): A-36
References:
NSCG RAI-10183-R1 Description of Change: An update to the PNPP LRA is required to address the PNPP corrective actions for the prevention of issues identified in NSCG RAI-10183-R1. The Open Cycle Cooling Water Systems Program requires enhancement to ensure the effectiveness of the Generic Letter (GL) 89-13 heat exchanger thermal performance monitoring.
PNPP LRA, Appendix A, Section A.1.38, Page A-36 is revised as follows:
A.1.37 OPEN-CYCLE COOLING WATER SYSTEM PROGRAM The open-cycle cooling water (OOCW) systems program is an existing program that manages material loss due to micro-or macro-organisms and various corrosion mechanisms to ensure effective transfer of heat from safety-related structures, systems and components (SSCs) to the ultimate heat sink (UHS). At PNPP, raw water for heat transfer to safety-related SSCs is accomplished with the emergency service water (ESW) system. The program relies on the implementation of the recommendations of NRC Generic Letter (GL) 89-13 to ensure that the effects of aging on the OCCW systems will be managed for the period of extended operation. Other components are also managed under OCCW based on exposure to a raw water environment and the aging management review. In accordance with the guidance of GL 89-13, the OCCW program manages aging effects by using a combination of preventive, condition, and performance monitoring activities. These include (a) surveillance and control techniques to manage aging effects caused by biofouling, corrosion, erosion, protective coating failures, and silting in the OCCW system or structures and components serviced by the OCCW system; (b) inspection of critical components for signs of corrosion, erosion, and biofouling; and (c) testing of the heat transfer capability of heat exchangers that remove heat from components important to safety.
Loss of material due to recurring internal corrosion has been identified for the Emergency Service Water (ESW) System. Loss of material due to recurring internal corrosion and erosion is managed by augmented inspections utilizing the XI.M17 Flow Accelerated Corrosion program. The XI.S7, RG 1.127, Inspection of Water Control Structures Associated with Nuclear Power Plants aging management program is credited with silt removal from the from the multi-port intake structure through the ESW pumphouse, including intake and alternate intake tunnels, associated tunnel riser shafts, discharge tunnel, and discharge structure. AMP XI.M41, Buried and Underground Piping and Tanks, manages aging effects for underground ESW piping. These activities are not managed by this program.
The program will be continued for the period of extended operation.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 3 Page 2 of 2 Enhancements The implementing procedures for heat exchanger thermal performance testing will be enhanced to require each heat exchanger thermal performance periodic test instruction to include the following steps (or similar) to evaluate the test results:
Provide the work order and planned date for the next scheduled test or cleaning for this heat exchanger.
Since the latest cleaning of this heat exchanger, if 2 or more valid heat exchanger test results are available, project the date for no margin to the acceptance criteria based on the current performance trend.
If the projected date for no margin will occur before the planned date for the next heat exchanger test or cleaning, initiate a Condition Report.
The enhancements will be implemented no later than six months prior to the period of extended operation.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 4 Page 1 of 1 Attachment No. 4 LRA Section: Appendix A, Section A.2.5.3 (as revised by L-24-108)
LRA Page Number(s): A-54
References:
TRP-063-03, Vistra Letters L-24-020 (LRA Supplement 2) and L-24-108 (LRA Supplement 3)
Description of Change: In the response to the referenced TRP in Vistra letters L-24-108 (Attachment 15) and L-24-020 (Attachments 42, 46, and 52), PNPP agreed to add the aging management of the containment piping penetrations bellows into the Fatigue Management Program. PNPP LRA Section A.2.5.3 (as revised by L-24-108) is revised to reflect the results of the fatigue analyses to incorporate the containment piping penetrations bellows into the PNPP Fatigue Management Program.
PNPP Section A.2.5.3, as previously modified by PNPP LRA Supplement 3 (Vistra Letter L-24-108), Page A-54, is revised as follows:
A.2.5.3 CONTAINMENT PIPING PENETRATIONS BELLOWS Guard pipe assemblies associated with containment penetrations utilize bellows. The PNPP specification required these bellows to be analyzed for at least 500 cycles of normal operation plus one safe shutdown earthquake (SSE) cycle for 40 years of operation.
Therefore, these fatigue analyses are identified as TLAAs requiring disposition for license renewal.
PNPP has evaluated the containment piping penetrations bellows fatigue and determined, based on the 40-year CUF for the bounding penetration bellows, that the bellows fatigue usage is bounded by the fatigue usage of the penetrations. Therefore, since the penetration fatigue is not expected to exceed allowable limits at 60-years neither would the bellows fatigue. However, since PNPPs method for managing fatigue is to track and evaluate transient cycles and to calculate CUFs to ensure that the CUFs for the limiting components do not exceed design limits, the effects of fatigue on the containment piping penetrations bellows will be managed for the period of extended operation by the Fatigue Monitoring Program. the containment piping penetration bellows that bound or represent all the other containment piping penetration bellows have been identified. The CUFs for these bounding containment piping penetration bellows will be re-evaluated based on actual plant transients to verify that the components remain within the ASME fatigue requirements through the period of extended operation.
The effects of fatigue on the containment piping penetrations bellows will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21(c)(1)(iii).
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 5 Page 1 of 3 LRA Section: Appendix A, Table A.3 LRA Page Number(s): A-73
References:
NCSG RAI-10183-R1 Description of Change: Recent PNPP corrective actions identified to prevent recurrence of issues identified in NSCG RAI-10183-R1. The Open Cycle Cooling Water Systems Program requires enhancement to ensure the effectiveness of Generic Letter (GL) 89-13 heat exchanger thermal performance monitoring.
PNPP LRA Appendix A, Table A.3, Page A-73 is revised as follows:
NOTE - Only the affected Table page and revised row are included. Unchanged pages of the Table are omitted.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 5 Page 2 of 3 Table A.3 License Renewal Commitments Item No.
AMP Commitment Implementation Schedule Related LRA Sections 33 XI.E2 Implement the new Non-EQ Instrumentation Circuits Program May 8, 2026 A.1.33 B.2.33 34 XI.E1 Complete the following enhancements to the existing Non-EQ Insulated Cables and Connections Program:
- 1. The program will be enhanced to include a plant-specific procedure for plant walkdowns of adverse localized environments.
May 8, 2026 A.1.34 B.2.34 35 XI.M32 Implement the new One-Time Inspection Program May 8, 2026 A.1.35 B.2.35 36 XI.M35 Implement the new One-Time Inspection of ASME Code Class 1 Small Bore Piping Program May 8, 2026 A.1.36 B.2.36 37 XI.M20 ContinueComplete the following enhancement to the existing Open Cycle Cooling Water System Program:
The implementing procedures for heat exchanger thermal performance testing will be enhanced to require each heat exchanger thermal performance periodic test instruction to include the following steps (or similar) to evaluate the test results:
Provide the work order and planned date for the next scheduled test or cleaning for this heat exchanger.
Since the latest cleaning of this heat exchanger, if 2 or more valid heat exchanger test results are available, project the date for no margin to the acceptance criteria based on the current performance trend.
OngoingMay 8, 2026 A.1.37 B.2.37
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 5 Page 3 of 3 37 (Cont.)
If the projected date for no margin will occur before the planned date for the next heat exchanger test or cleaning, initiate a Condition Report.
38 XI.S8 Complete the following enhancement to the existing Protective Coating Monitoring and Maintenance Program:
- 1. The existing PNPP Protective Coating Monitoring and Maintenance Program will be enhanced to comply with the requirements of ASTM D5163-08.
May 8, 2026 A.1.38 B.2.38
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 6 Page 1 of 2 LRA Section: Appendix B, Table B.1-2 LRA Page Number(s): B-13
References:
NCSG RAI-10183-R1 Description of Change: Recent PNPP corrective actions have identified an adverse trend regarding thermal performance monitoring of Generic Letter (GL)-89-13 heat exchangers. The Open Cycle Cooling Water Systems Program requires enhancement to ensure the effectiveness of GL 89-13 heat exchanger thermal performance monitoring.
PNPP LRA, Table B.1-2, Page B-13 is revised as follows:
NOTE - Only the affected Table page and revised row are included. Unchanged pages of the Table are omitted.
Table B.1-2 Consistency of PNPP Aging Management Programs with NUREG-1801 Program Name New/
Existing Consistent with NUREG-1801 Consistent with NUREG-1801 with Exceptions Plant Specific Enhancement Required Neutron-Absorbing Materials Other Than Boraflex Existing Yes Non-EQ Electrical Cable Connections Program New Yes Non-EQ Inaccessible Power Cables Program Existing Yes Yes Non-EQ Instrumentation Circuits Program New Yes Non-EQ Insulated Cables and Connections Program Existing Yes Yes One-Time Inspection Program New Yes
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 6 Page 2 of 2 Program Name New/
Existing Consistent with NUREG-1801 Consistent with NUREG-1801 with Exceptions Plant Specific Enhancement Required One-Time Inspection of ASME Code Class 1 Small Bore-Piping Program New Yes Open-Cycle Cooling Water System Program Existing Yes
--Yes Protective Coating Monitoring and Maintenance Program Existing Yes Yes Reactor Head Closure Stud Bolting Program Existing Yes Yes Yes Reactor Vessel Surveillance Program Existing Yes RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program Existing Yes Yes
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 7 Page 1 of 6 LRA Section: Appendix B, Section B.2.37 LRA Page Number(s): B-103 through B-107
References:
NCSG RAI-10183-R1 Description of Change: An update to the PNPP LRA is required to address the PNPP corrective actions for the prevention of issues identified in NSCG RAI-10183-R1. The Open Cycle Cooling Water Systems Program requires enhancement to ensure the effectiveness of the Generic Letter (GL) 89-13 heat exchanger thermal performance monitoring. Also, an editorial change is made to change the text degraded condition to loss of system intended function to align with the source document.
PNPP LRA, Appendix B, Section B.2.37, Page B-103 through B-107 are revised as follows:
B.2.37 OPEN-CYCLE COOLING WATER SYSTEM PROGRAM Program Description The open-cycle cooling water (OCCW) systems program is an existing program that manages material loss due to micro-or macro-organisms and various corrosion mechanisms to ensure effective transfer of heat from safety-related structures, systems and components (SSCs) to the ultimate heat sink (UHS). At PNPP, raw water for heat transfer to safety related SSCs is accomplished with the emergency service water (ESW) system. The program relies on the implementation of the recommendations of the Nuclear Regulatory Commission (NRC) Generic Letter (GL) 89-13 to ensure that the effects of aging on the OCCW systems will be managed for the period of extended operation. In accordance with the guidance of GL 89-13, other components are also managed under OCCW based on exposure to a raw water environment and the aging management review. The OCCW program manages aging affects by using a combination of preventive, condition, and performance monitoring activities. These actions include (a) surveillance and control techniques to manage aging effects caused by biofouling, corrosion, erosion, protective coating failures, and silting in the OCCW system or structures and components serviced by the OCCW system; (b) inspection of critical components for signs of corrosion, erosion, and biofouling; and (c) testing of the heat transfer capability of heat exchangers that remove heat from components important to safety. AMP XI.M17, Flow-Accelerated Corrosion, manages the aging effects caused by pipe and piping component wall thinning.
AMP XI.M21A, Closed Treated Water Systems, manages closed cooling water systems.
Loss of material due to recurring internal corrosion has been identified for the Emergency Service Water (ESW) system. Loss of material due to recurring internal corrosion and erosion is managed by augmented inspections utilizing the XI.M17, Flow Accelerated Corrosion program. The XI.S7 RG 1.127, Inspection of Water Control Structures Associated with Nuclear Power Plant aging management program is credited with silt removal from the multi-port intake structure through the ESW pumphouse, including intake and alternate intake tunnels, associate tunnel riser shafts, discharge tunnel, and discharge structure.
AMP XI.M21A, Closed Treated Water Systems, manages closed cooling water systems. AMP
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 7 Page 2 of 6 XI.M41, Buried and Underground Piping and Tanks, manages aging effects for underground ESW piping. These activities are not managed by this program.
ESW system components are unlined. The recommendations of LR-ISG-2013-01 to include recommendations of LR-ISG-2013-01 section XI.M42, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks do not apply.
The program will be continued for the period of extended operation.
NUREG-1801 Consistency The open-cycle cooling water (OCCW) systems program is an existing PNPP program that, with enhancement, will be is consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.M20, Open-Cycle Cooling Water and additional guidance in LRISG-2012-02.
Exceptions to NUREG-1801 None Enhancements NoneThe program will be enhanced as follows:
The implementing procedures for heat exchanger thermal performance testing will be enhanced to require each heat exchanger thermal performance periodic test instruction to include the following steps (or similar) to evaluate the test results: Program Element Affected: Monitoring and Trending (Element 5)
Provide the work order and planned date for the next scheduled test or cleaning for this heat exchanger.
Since the latest cleaning of this heat exchanger, if 2 or more valid heat exchanger test results are available, project the date for no margin to the acceptance criteria based on the current performance trend.
If the projected date for no margin will occur before the planned date for the next heat exchanger test or cleaning, initiate a Condition Report.
The enhancements will be implemented no later than six months prior to the period of extended operation.
Operating Experience The following operating experience examples provide objective evidence that the Open-Cycle Cooling Water Systems Program will be effective in ensuring that component intended functions are maintained consistent with the current licensing basis during the period of extended operation.
A review of plant specific PNPP operating experience since 2013, was conducted through a search of plant corrective action program documents identified condition reports (CRs) potentially identifying operating experience related to the OCCW aging management
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 7 Page 3 of 6 program. The review shows the guidance of NRC GL 89-13 has been effective in managing aging effects due to biofouling, corrosion, erosion, protective coating failures, and silting in structures and components serviced by OCCW systems. The PNPP corrective action program identified more recent adverse trends regarding thermal performance monitoring of GL 89-13 heat exchangers. An enhancement to this aging management program, combined with actions from the corrective action program, are expected to ensure the effectiveness of GL 89-13 heat exchanger thermal performance monitoring going forward.
Annual diving inspections have effectively ensured the intake and discharge structures continue to control mussel contamination within acceptable limits. No evidence of Asian clams has been identified. There have been no documented instances of mussel infestation in ESW system piping or safety-related heat exchangers served by the system. The lack of documented instances indicates that the existing program has been effective in managing aging effects due to biofouling and silting in structures and components serviced by OCCW systems.
Periodic heat exchanger performance tests ensure the effective transfer of heat from safety-related structures, systems and components (SSCs) to the ultimate heat sink (UHS).
No unacceptable performance has been documented.
In April 2013, a Condition Report (CR) documented that the ECC B heat exchanger divider plate was found to have a wall thickness that appeared to be less than allowable. Extent of condition found the ECC A heat exchanger acceptable. An engineering review of calculations and divider plate condition concluded there was some wall thinning that was acceptable for the period leading to the next scheduled inspection.
A January 2022 CR identified a degrading trend in the Div 1 RHR B train heat exchanger (HX) thermal performance. The margin between calculated HX performance and acceptance criteria was identified as slim. Subsequent analysis of another segment of test data from the same performance test identified additional performance margin. A subsequent performance test for the same heat exchanger in June 2022 indicated adequate thermal performance.
A separate CR written in January 2022 identifying a degrading trend on the RHR A train heat exchangers thermal performance. A heat exchanger thermal performance test was subsequently scheduled and performed in November 2022 that indicated adequate heat exchanger thermal performance, still with low margin. A work order was scheduled to clean the RHR A train heat exchangers that was completed in April 2023 restoring margin.
In February 2022, a CR documented that of a heat sink self-assessment identified that the Division 1 diesel generator jacket water heat exchanger had a degrading trend in performance and as a result, corrective actions were established to ensure cleaning and restoration of margin in the heat transfer coefficient.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 7 Page 4 of 6 A February 2022 CR identified a degrading trend in the Div 3 EDG jacket water HX. The margin between calculated HX performance and acceptance criteria was identified as slim. The HX was cleaned in May 2022 and subsequent thermal performance testing results were acceptable.
In February 2022, a rollup CR was initiated to document that four separate Generic Letter 89-13 heat exchangers were exhibiting degrading thermal performance trends but neither a CR or a maintenance notification was initiated at the time of the testing to document the condition. The rollup CR identified recent CRs that captured those issues. Significantly, the rollup CR extent of condition review included statements that were found to be inaccurate, such that a new CR was initiated in September 2024 to have the statements corrected.
A March 2022 CR identified the Emergency Closed Cooling (ECC) A HX test conditions, when projected to design basis conditions, the test acceptance criteria could not be met. Review found there was insufficient heat load to obtain valid test results. Several related human performance errors were identified with the test and documented in a subsequent CR. The HX performance test was reperformed in June 2022. Analysis of the June test results resulted in a third CR. Excess conservatism had been included in the performance acceptance criteria (additional heat load and tube plugging margin). When the excess conservatism was removed, the test results were considered acceptable, but with low margin. The HX was cleaned in the following year to restore margin.
A May 2022 CR identified a degrading trend in the Div 3 EDG jacket water HX was found unsatisfactory when opened and inspected. A reddish-brown substance, initially identified as zebra mussel shells but later determined to be carbon steel corrosion products, had 72 tubes more than 50% blocked. The tubes and waterbox were cleaned and cleared.
In March 2023 two CRs documented that fouling in RHR heat exchangers was observed during cleaning, demonstrating that, while fouling occurs, site practices are effective at identifying and controlling the fouling before unacceptable heat exchanger performance occurs.
AMP XI.M17, Flow-Accelerated Corrosion, manages the aging effects caused by pipe and piping component wall thinning and corrosion. This includes identification of internal erosion and corrosion of valve bodies, piping, piping components and piping elements. A review of the OE potentially related to the XI.M20 OCCW aging management program and of the OE potentially related to the XI.M17 program was conducted. Operating experience with the applicable principal aging mechanisms are identified below.
Through wall leaks - aging mechanism internal corrosion Through wall, or pinhole, leaks are usually the result of corrosion over time, at unpredictable locations in a piping system. No evidence of recurring leaks in a single location have been found, though a very few occasions have noted an increased leak rate at a known leak location prior to repair. None of the leaks resulted in degraded performance loss of system function. Numerous instances
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 7 Page 5 of 6 have been documented of pipe leaks or blockage due to corrosion between 2013 and 2022 related to XI.M20 or XI.M17.
In August 2017 a CR documents that a leak was identified in ESW piping near ECC heat exchanger B. An evaluation determined the leak was a pinhole leak rather than cracking. The extent of the leak did not affect operability. A clamp was applied to contain the leak until repairs could be accomplished. Another leak in the vicinity prompted an extent of condition evaluation which found the other areas to be satisfactory. This section of pipe was removed and eliminated by design in Refuel Outage (RFO) 17 so it could not happen again, and the same piece in the A loop was eliminated in RFO 18.
The XI.M17 walkdown process is effective in identifying leaks and the management and repair of those leaks have demonstrated the program will be effective throughout the period of extended operation. The XI.M20 program does not need to be enhanced for this aging effect.
Wall thinning - aging mechanism internal erosion.
Wall thinning occurs as a result of erosion by corrosion products suspended in the fluid, high velocity or high pressure flow. Thinning occurs most often at points where the piping system turns or where piping opens into a waterbox of some kind, impacting a localized surface within the waterbox. Numerous instances have been documented of wall thinning or flange or valve seat leaks due to erosion from 2013 through 2021.
In April 2018 a CR documented that a through-wall pipe leak was discovered on ESW 'A' piping at the outlet of the RHR 'A' heat exchangers. The leak was identified between valves 1P45F550A (RHR A/C HX'S ESW OUTLET) and 1P45F0068A (RHR A HX'S ESW OUTLET VALVE) on 20" piping beneath piping insulation. The leak did not affect operability.
In July 2018 a CR documented that a UT examination as part of the FAC program extent of condition evaluation for a condition report found localized wall thinning.
While the amount of wall thinning was acceptable, a 30-day UT monitoring plan was instituted to determine the rate of thinning. Replacement of the elbow occurred in the next refueling outage.
The XI.M17 ultrasonic examination process is effective in identifying wall thinning and management or repair of the condition to ensure no loss of function. These actions have demonstrated the program will be effective throughout the period of extended operation. The XI.M20 program does not need to be enhanced for this aging effect.
Other aging mechanisms, such as external corrosion, have been isolated instances that were identified in system walkdowns and tracked or repaired using the corrective action program.
Perry Nuclear Power Plant LRA Supplement 5 L-24-179 Attachment 7 Page 6 of 6 The XI.M20 program, in conjunction with the XI.M17 and XI.M41 programs have been effective in managing aging effects due to biofouling, corrosion, erosion, protective coating failures, and silting in structures and components serviced by OCCW systems.
Conclusion The Open-Cycle Cooling Water System program will provide reasonable assurance that aging effects will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis during the period of extended operation.