L-25-013, License Renewal Application, Revision O - Supplement 8
| ML25027A327 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/27/2025 |
| From: | Penfield R Vistra Operations Company |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-25-013 | |
| Download: ML25027A327 (1) | |
Text
L-25-013 January 27, 2025 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 8
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) 6555 SIERRA DRIVE IRVING, TEXAS 75039 o 214-812-4600 VISTRACORP.COM
Perry Nuclear Power Plant L-25-013 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-24-256 from Rod L. Penfield to the Nuclear Regulatory Commission, dated December 19, 2024, submitting the Perry Nuclear Power Plant License Renewal Application Revision 0, Supplement 7 (ADAMS Accession No. ML24354A265)
- 12. 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)
- 13. 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)
- 14. Letter L-24-209, from Rod L. Penfield to the Nuclear Regulatory Commission, dated November 19, 2024, submitting the License Renewal Application for the Perry Nuclear Power Plant - Response to Request for Additional Information - Set 3 (ADAMS Accession No.
M L24324A 185)
- 15. 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. ML24305A134)
- 16. Letter L-24-257, from Rod L. Penfield to the Nuclear Regulatory Commission, dated December 4, 2024, submitting the License Renewal Application for the Perry Nuclear Power Plant Revision 0 - Response to Requests for Confirmatory Information - Set 2 (ADAMS Accession No. ML24339A066) 6555 SIERRA DRIVE IRVING, TEXAS 75039 o 214-812-4600 VISTRACORP.COM
Perry Nuclear Power Plant L-25-013 Page 3 of 3 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 11 ).
In addition, as a result of the NRC's review and audit of the PNPP LRA, the NRC Staff has submitted and the PNPP Staff responded to three sets of Requests for Additional Information (RAls) (References 12 through 14) and two sets of Requests for Confirmatory Information (RC ls) (References 15 and 16).
The attachment to this letter provides Supplement 8 of the PNPP LRA, which provides the PNPP LRA updates previously addressed in the response to the NRC's second set of RAls (Reference 13).
The regulatory commitments identified in Appendix A, Table A.3 of the PNPP LRA are not impacted by this LRA supplement. 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 January 27, 2025.
Sincerely,
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Attachment:
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 0
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachments Index Page 1 of 2 PNPP LRA Supplement 8 Attachments Index for Safety Relief Valve (SRV) Flex Hose Issue Attachment No.
LRA Section, Table or Appendix Supplemented Subject Source 1
Table 3.1.2-2 Update the response to NCSG RAI-10276-R1 to reflect results of the condition report investigation related to SRV flexible hose issues.
Response to RAI NCSG RAI-10276-R1 (Vistra Letter L-24-208, ), and the associated PNPP Corrective Action Program (CAP) investigation 2
Standard Notes for Tables 3.1.2-1 through Table 3.1.2-6 Update the standard note to the table to reflect results of the condition report investigation related to SRV flexible hoses. and the Response to RAI NCSG RAI-10276-R1 (Vistra Letter L 208, Attachment 2) 3 Table 3.1.1, Item 3.3.3-107 Update the table to address the External Surfaces Monitoring of Mechanical Components program that will manage the aging effects of SRV flexible hoses. and the Response to RAI NCSG RAI-10276-R1 (Vistra Letter L 208, Attachment 2) 4 Appendix A Section A.1.18 Incorporate additional details regarding SRV flexible hoses to the External Surfaces Monitoring of Mechanical Components Program description. and the Response to RAI NCSG RAI-10276-R1 (Vistra Letter L 208, Attachment 2) 5 Appendix B Section B.2.18 Incorporate additional details regarding SRV flexible hoses to the External Surfaces Monitoring of Mechanical Components Program description. and the Response to RAI NCSG RAI-10276-R1 (Vistra Letter L 208, Attachment 2)
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 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 the Request for Additional Information (RAI) responses which were submitted via the following Vistra correspondence:
In addition to the above supplements, the annual update and the RAI responses, the following letters provided responses to the two sets of Requests for Confirmatory Information (RCIs) received:
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 in the attachments, blue text is retained from the original application hyperlinks for consistency and so this text does not represent any change to LRA content. Also 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 8 L-25-013 Attachment 1 Page 1 of 20 LRA Section: Section 3.1.2.1.2, Table 3.1.2-2 LRA Page Number(s): Pages 3.1-2 and 3.1-3, Pages 3.1-47 thru 3.1-55
References:
NCSG RAI-10276-R1, Perry Nuclear Power Plant (PNPP) Responses to RAIs Round 1 (Set 2), Vistra Letter L-24-208, Attachment 2 Description of Changes: In the response to the NRC Request for Additional Information (RAI)
NCSG RAI-10276-R1 in Vistra Letter L-24-208 dated October 2, 2024 (Attachment 2), a future supplement to the PNPP License Renewal Application (LRA) was to be submitted after the completion of the investigation of Condition Report (CR) 2024-07520. The RAI topic involved PNPP operating experience with stainless steel flex hose failures. This LRA supplement provides the results of the CR investigation that addresses the likely cause of the stainless steel flex hose failures, including the source of the chloride contamination found on two failed Safety Relief Valve (SRV) flex hoses and the contributing conditions to the flex hose failures. This LRA supplement also provides updates to the PNPP LRA to reflect the planned actions and conclusions of the completed CR investigation. NCSG RAI-10276-R1 referred to several SRV flex hose failures from PNPPs internal operating experience. The following list summarizes that operating history:
2011 - 1B21F0051B Air Hose -The instrument air supply line (hose assembly) was found to be leaking at the bottom of the metal hose where the braided jacket meets the ferrule.
2017 - 1B21F0041B - The leak detailed in the CR occurred on a bent area of the hose.
This is the first failure of the 41B in the original hose.
2021 - 1B21F0041B - The crack was approximately the length of 1/2 of an inch on an inner corrugation of the metal hose.
2021 - 1B21F0051B - The supply flex hose was snooped, and bubbles indicated that the hose was leaking.
2023 - 1B21F0041B - Inspection revealed that braided air line was damaged.
2023 - 1B21F0047B - During performance of work to replace a leaking flex hose on the air supply piping to SRV 1B21F0041B, the workers observed that the flex hose for the air supply piping to SRV 1B21F0047B had also developed an air leak.
As part of the investigation from CR 2024-07520, the failure mechanisms were evaluated for the two SRV flex hose failures identified in 2023. The subject hoses had notably different service lives with the first hose (being original plant equipment) having failed after approximately 36 years of service while the second failure (being of a different design) occurred in approximately 2 years of service. A difference in the number of bellows convolutions exists between the two hose designs and this difference affected the ability of each flex hose to accommodate bending incurred during flex hose installation.
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The higher number of bellows convolutions in the longer-lived hose allowed it to accommodate bending during installation with minimal or no immediate microcracking. However, the overall appearance of the crack network in the cross section of this hose is consistent with classic stress corrosion cracking (SCC).
The lower number of bellows convolutions in the shorter-lived hose resulted in more stress from bending during installation, which drove crack initiation and propagation much faster than that seen in the other hose failure. This bending crack, or ductile overload failure, of the bellows did not require chlorides to initiate and propagate. Subsequent exposure of the crack to the chloride containing leak detector solution, coupled with thermal and vibration cycles, may have played a role in any final fracture of the bellows base metal that had not broken through-wall after initial installation.
Corrective actions within the sites Corrective Action Program (CAP) have been generated. The planned corrective actions included a design change to the hose design assembly to reduce the potential for excessive installation stresses, enhanced procedural guidance for hose installation, revised procedural guidance for leakage testing, and the periodic replacement of the hoses. The initial replacement frequency will be every 3 operating cycles.
Source of flex hose chloride contamination:
PNPP maintains a chemical control program that requires all materials meeting the definition of a chemical to be approved for the intended use prior to purchase or receiving the material on-site. Evaluations of chemicals intended for use in or on critical use areas for secondary plant systems, or safety related equipment are performed to determine system contaminant concerns and assure chemicals are approved for the intended application. The references for this program include the General Electric BWR Operators Manual for Materials and Processes, often referred to as the Red Book. The Red Book includes requirements for leak detector solution and provides allowable limits for contaminants, including chlorine.
As noted in the CR investigation, the same leak detector solution has been used at PNPP for leakage inspections, including the subject flex hoses. The investigation determined that leak detector solution, applied through the wire braid of a flex hose and onto the bellows, leaves a residue between the bellows convolutions as the leak detector solution evaporates. Constituents of the leak detector solution, notably chlorine, was identified in some of the flex hose cracks examined after the 2023 flex hose failures. Subsequent leak detector solution applications result in concentrating deposits, potentially capable of yielding a higher-than-expected concentration of certain contaminants, including chlorides. Compared to other components, flex hose bellows with braided sheathing uniquely retain leak detector solution residue.
It should be noted that the Red Book also indicates that leak detector solution should be removed with demineralized water, a practice that is not currently implemented at the site. It is expected that this practice would mitigate the accumulation of leak detector solution residue in the corrugations of the subject flex hoses.
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Given the above, a credible source of chloride contamination of the flex hose bellows convolutions has been established. Importantly, an action to remove leak detector solution with demineralized water after leak testing would mitigate the source of flex hose chloride contamination. Actions within the CAP will address updating the appropriate testing procedure(s).
Contributing conditions to the flex hose failures:
The 2023 failure analysis report for two of the subject flex hoses attributed the failure cause to chlorine induced stress corrosion cracking. However, there are contributing conditions that warrant consideration since they would also influence the likelihood of stress corrosion cracking in these flex hoses.
The manufacturers drawings of the flex hoses prescribe a geometric configuration of the installed flex hose, but do not cite restrictions against bending the flex hose beyond those dimensions during installation. Exceeding the minimum flex hose bend radius significantly increases the stress in the bellows. Thus, a configuration change is planned to add a 45-degree fitting that would allow an increase in the overall bend radius of the flex hose and reduce the overall physical bending, significantly reducing the bending stress. Corrective actions within the PNPP CAP will address the change to the hose design.
In addition to the initial installation of SRV flex hoses, the hoses are subsequently disconnected and re-connected as part of periodic SRV replacements. Thus, the installation activity for these hoses is recurrent, and there is potential that the hoses could have been subject to inadvertent excessive bending during any of those activities.
The CR investigation concluded that cracks were likely initiated in the flex hose bellows as the result of excessive bending during installation and subsequent handling of the flex hoses. Both flex hoses analyzed in 2023 leaked in the same region where the most aggressive bending occurs along the outer surface of the bend of the lower connection. Differences in hose design between the two specimens tested after the 2023 failures likely explain the difference in service life. Specifically, the original plant equipment hoses were manufactured with six convolutions per inch resulting in greater tolerance to initial installation stresses. Conversely, the replacement hoses (such as the short-lived failure identified in 2023) were manufactured with only four convolutions per inch. The difference in hose design exacerbated the effects of bending during installation and subsequent handling.
A 2022 engineering change that was never implemented (prior to the 2023 failures and resulting laboratory testing/analysis) cited industry operating experience of a hose failure attributed to fatigue caused by the hose being moved off its XY plane, installation mishandling, and stress corrosion cracking.
Based on these potential causes and uncertainties, the planned corrective actions included a design change to the hose design assembly to reduce the potential for excessive installation stresses, enhanced procedural guidance for hose installation, revised procedural guidance for
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 1 Page 4 of 20
leakage testing, and the periodic replacement of the hoses. The initial replacement frequency will be every 3 operating cycles.
With updated instructions for the proper handling and installation of the flex hoses, and leak testing that mitigates exposures to chlorides, the combined actions are considered to restore SRV flex hose reliability. A design change to add a 45-degree fitting to increase the overall bend radius of the flex hose would further improve design margin. Also, requiring periodic replacement of the flex hoses will ensure that any potential failure uncertainties will not affect continued reliability.
Justification of 3-cycle frequency for replacement of the flexible hoses is as follows.
The subject hoses provide compressed air to Safety/Relief Valve (SRV) actuators. PNPPs design includes 19 such SRVs, each with one flexible hose in the compressed air supply. Of these, eight are designated as Automatic Depressurization System (ADS) SRVs. Compressed air is provided to the ADS SRVs by the Safety Related Instrument Air System (system designation P57). Compressed air is provided to the non-ADS SRVs by the Instrument Air System (system designation P52). All 19 of the subject hoses are tested at least every other refueling outage under PNPPs In-Service Inspection (ISI) Program. The eight ADS SRV hoses are also subject to a P57 System air leakage test performed every refueling outage. Only 3 out of 19 SRVs have experienced hose failures. Of these, only valve 1B21F0041B has experienced failures that occurred in a timeframe less than the proposed replacement frequency. Valve 1B21F0041B is the only ADS SRV of the three SRVs in the failure population. As an ADS SRV, this valve would be subject to the system leakage test performed each refueling outage. All three are subject to ISI leak detection testing at least every other outage. The failures investigated in detail in CR 2024-07520 have been shown to be event-based mechanisms rather than aging effects; specifically, installation-induced stresses that resulted in significantly reduced service life. The actions prescribed by CR 2024-07520 will limit exposure of the hoses to excess stresses from mishandling, reduce or eliminate exposure to chlorides and replace the hoses with an enhanced design. With those actions implemented, there is no need to periodically replace the hoses, other than to ensure that any uncertainties affecting long term reliability are bound. The reason for a 6-year replacement interval is that it is considered very conservative compared to 40 years, and sufficient to bound uncertainties not identified by the evaluations of hose failures to date. The 3-cycle initial replacement frequency is therefore judged to be appropriate based on operating history and existing testing which can identify pressure boundary loss between replacement activities.
Updates to NCSG RAI-10276-R1 indicated actions:
In the PNPP response to the RAI, the following action was identified as recommended within the Condition Report text:
Although the Drywell environment is not supplied with outdoor air, the presence of chloride contamination on hoses originally installed at PNPP and subsequent SCC failures,
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warrants further evaluation. The first two bulleted recommended actions in CR-2024-07520 above will address these issues. Those actions are:
A population of the flex hoses available in the warehouse should be analyzed (destructive testing) to determine if the presence of chlorides, determined previously to be causal to hose failure, is the result of a manufacturing issue.
In-plant sampling should be performed in the area of the SRVs to definitively conclude that the chloride source is not plant-specific.
Neither of the above actions is considered to be warranted following the above additional investigation. A credible source of chlorides on the flex hoses has been identified, and discussions with the manufacturer of the flex hoses affirmed that chloride contamination during the manufacture and handling of the flex hoses is highly improbable based on their operating experience.
In-plant sampling in the Drywell can only be performed during a refueling outage. However, based on the conclusions of the evaluation performed in the Condition Report, a specific and highly localized source of chloride contamination has been identified.
LRA Changes In the PNPP response to the RAI, the following changes to LRA Table 3.1.2-2, Nuclear Boiler System were proposed, pending the results of the follow-up investigation:
Rows 10 and 11 will be removed or replaced with a different material, if appropriate.
These rows reflect a design change to accommodate a nickel alloy flex hose material that will not be implemented.
Plant Specific Note 111 on LRA Page 3.1-110 is no longer applicable since Row 10 is being removed or replaced with an improved material. The note will be deleted or revised accordingly.
A new line item similar to Row 13 will be added replacing None with an aging effect of Cracking (due to SCC) and the aging management program with the External Surfaces Monitoring of Mechanical Components. The same NUREG-1801 Item and Table 1 Item will be listed. Standard Note F and a plant specific note assigned to clarify the limitations to this component based upon the conclusions regarding the extent of chloride contamination.
Based on corrective actions to ensure component reliability, these flexible hoses will be periodically replaced early in the period of extended operation (PEO). PNPP is conservatively retaining these flexible hoses in Table 3.1.2-2 until all of the originally installed flexible hoses are replaced. The three bullets above will be revised to:
Rows 10 and 11 will be removed. These rows reflect a design change to accommodate a nickel alloy flex hose material that subsequently was not, and will not be, implemented.
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Plant Specific Note 111 on LRA Page 3.1-110 is no longer applicable since Row 10 is removed. Note 111 will be revised as described below.
Row 12 is renumbered as Row 10 and represents the internal surfaces of the stainless steel flexible hoses associated with the air supplies to safety relief valves.
A new line item (numbered as Row 11) similar to Row 13 will be added replacing None with an aging effect of Cracking (due to various causes such as installation-initiated cracking exacerbated by chloride induced SCC, chloride induced SCC, and / or cycle fatigue) and the External Surfaces Monitoring of Mechanical Components program. This component type refers to the stainless steel flexible hoses associated with the air supplies to safety relief valves. The same NUREG-1801 Item and Table 1 Item will be listed for clarity. Since the aging effect is not in NUREG-1801 for this component, material, and environment combination, Standard Note H, and revised plant specific Note 111 will be assigned. Revised Note 111 will clarify that these components will be replaced initially at a 3-cycle frequency and will extend the service lives into the PEO.
Rows 13 and 14 (renumbered as Rows 12 and 13) refer to the stainless steel flexible hoses in the SRV stem leak off piping that discharges below the suppression pool surface. Cracking has not been identified in these flexible hoses.
Supplement 8, See Attachment 2 The text in the existing plant specific Note 111 will be deleted. The revised Note 111 will read:
NUREG-1801 provides no aging effects for this component type, material, and environment combination. PNPP assigns Cracking due to various causes such as installation-initiated cracking exacerbated by chloride induced SCC, chloride induced SCC, and / or cycle fatigue is assigned to this row. Two flexible hoses are associated with each Safety Relief Valve (SRV). This row represents the external surface of the flexible hoses supplying compressed air (Air - dry (Int) environment) to the operator of each Safety Relief Valve (SRV) in the main steam lines. These flexible hoses will be replaced initially at a 3-cycle frequency and will extend the service lives into the PEO.
Supplement 8, See Attachment 3 In PNPPs response to the RAI the following change to LRA Table 3.1.1, Line 3.1.1-107 Discussion Text was proposed:
Consistent with NUREG-1801, with the following clarification and a different program, the External Surfaces Monitoring of Mechanical Components program will manage cracking due to stress corrosion cracking of stainless steel flexible hoses suppling compressed air to the safety relief valves in the Nuclear Boiler system. See LRA Appendix B, B.2.18 for the associated commitments. In addition to the Reactor Vessel, Internals, and Reactor Coolant systems; stainless steel commodities in concrete in the Bulk Civil Commodities, Containment Structure, and
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Turbine Buildings and Associated Structures, Process Facilities, and Yard Structures have been aligned with this item.
Based on the action to periodically replace the stainless steel flexible hoses supplying compressed air to the safety relief valves in the Nuclear Boiler system, the above change to LRA Table 3.1.1, Line 3.1.1-107 Discussion Text will be revised to:
Consistent with NUREG-1801, with the following clarification and a different program. The External Surfaces Monitoring of Mechanical Components program will manage cracking due to various mechanisms of stainless steel flexible hoses supplying compressed air to safety relief valves in the Nuclear Boiler system. See Appendix B, Section B.2.18. In addition to the Reactor Vessel, Internals, and Reactor Coolant systems; stainless steel commodities in concrete in the Bulk Civil Commodities, Containment Structure, and Turbine Buildings and Associated Structures, Process Facilities, and Yard Structures have been aligned with this item.
Supplement 8, See Attachments 4 and 5 In PNPPs response to the RAI the following changes to LRA Appendix A, Section A.1.18 and Appendix B, B.2.18 were proposed:
In a future supplement to the LRA to be submitted after completion of CR-2024-07520 investigation, the following two changes are proposed:
Based on the completed Condition Report 2024-07520 investigation and identified corrective actions, a commitment will be added to LRA Sections A.1.18and B.2.18 that address actions to be completed before entry into the PEO regarding determining the extent of chloride contamination in the PNPP Drywell.
Based on the results of those actions, the aging effect of Cracking will be added to AMR tables for stainless steel components in an indoor air (uncontrolled) environment in the Drywell, if the aging affect is found to apply, to be managed by the External Surfaces Monitoring of Mechanical Components aging management program.
The operating experience section of LRA Appendix B, Section B.2.18, External Surfaces Monitoring of Mechanical Components, will be revised to include a discussion involving Condition Report 2024-01530 and related, recent Condition Reports 2024-07520 and 2024-07527.
Based on the investigation performed for CR 2024-07520, a source of chloride contamination and material susceptibility was confirmed that would be unique to these stainless steel flexible hoses. The repetitive application of leak detector solution on these components promotes concentration of contaminates within the flexible hose corrugations. Thus, the changes cited in the first bullet above are no longer warranted. Since the chloride contamination source is highly localized, and there is no evidence of widespread chloride exposure in the PNPP Drywell
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(evidence of stainless steel pitting or corrosion), the extent of chloride contamination in the PNPP Drywell has been determined. Based on the highly localized nature of the chloride contamination, generically adding the aging effect of cracking to AMR tables for stainless steel components in an indoor air (uncontrolled) environment in the Drywell is not warranted. Going forward, the process change in leak testing of flexible hoses supplying compressed air to the SRVs is expected to mitigate the concentration of chloride contamination on the SRV flex hoses.
The following paragraph will be added to the description sections of LRA Section A.1.18, which will be part of the PNPP LRA UFSAR Supplement and LRA Section B.1.18.
PNPP does not have an air environment containing halides. Stainless steel flexible hoses supplying compressed air to the main steam line safety relief valves located in the Drywell have experienced cracking on the external surface. These surfaces, typically exposed to air, have been periodically exposed to a local source of chloride contaminates from a leak detector solution. The external surface of the pressure boundary for these flexible hoses is physically inaccessible. It is enclosed in a stainless steel mesh integral to its structural integrity. Consequently, these components will be managed by periodically replacing them at a frequency that provides reasonable assurance the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.
The change to the operating experience section of LRA Appendix B, Section B.2.18, External Surfaces Monitoring of Mechanical Components for this issue will include the following:
In 2021, condition reports document leaking stainless steel flexible air supply hoses for safety relief valves 1B21F0041B and 1B21F0051B. The condition reports cite the suspected failure causes as cyclic fatigue or cracking due to stress corrosion cracking. Both condition reports note that the causes of leaks are not yet known and that a failure analysis is needed to definitively determine the cause. In lieu of a failure analysis, PNPP developed an equivalent replacement hose change document with an improved material resistant to stress corrosion cracking. However, this replacement modification was not implemented. Although failure analyses were ultimately not performed on the flex hoses for the leaks identified in 2021, in 2023 new condition reports identified additional leaking stainless steel flexible air hoses for safety relief valves 1B21F0041B (replaced in 2021) and 1B21F0047B (original equipment). PNPPs operating experience review performed for these condition reports identified that previous leaks in the flexible air hoses for the same system had occurred in 2021 (SRV-0041B and -0051B), in 2017 (SRV-0041B), and in 2011 (SRV-0051B). An outside vendor performed a failure analysis for the leaking flex hoses from 2023, and their failure analysis reports determined the cause to be outside diameter chloride induced stress corrosion cracking. In 2024 an additional condition report was initiated to determine the source of the chloride contamination of the flex hoses, and to prescribe corrective actions to ensure component reliability. The source of the chloride contamination was attributed to accumulated residue from leak detector solution in the corrugations of the flex hoses under the stainless steel braided sheathing. The planned corrective actions
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included a design change to the hose design to reduce the potential for excessive installation stresses, enhanced procedural guidance for hose installation, revised procedural guidance for leakage testing, and the periodic replacement of the hoses. The initial replacement frequency will be every 3 operating cycles.
PNPP Section 3.1.2.1.2, Pages 3.1-2 and 3.1-3, as supplemented by Supplement 2,, Vistra Letter L-24-020 (pdf page 105 and 106) is revised as follows: (see following pages)
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3.1.2.1.2 Nuclear Boiler Materials Nuclear Boiler components are constructed of the following materials:
aluminum copper alloy >15% Zn nickel alloy stainless steel steel Environments Nuclear Boiler components are exposed to the following environments:
air - dry air - indoor, uncontrolled treated water treated water >60°C (>140°F)
Aging Effects Requiring Management The following aging effects associated with the Nuclear Boiler components require management:
cracking cumulative fatigue damage loss of material loss of preload Aging Management Programs The following aging management programs manage the effects of aging on Nuclear Boiler components:
Bolting Integrity (B.2.7)
Compressed Air Monitoring (B.2.16)
External Surfaces Monitoring of Mechanical Components (B.2.18)
Flow-Accelerated Corrosion (B.2.22)
Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.25)
One-Time Inspection (B.2.35)
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Selective Leaching (B.2.42)
TLAA Water Chemistry (B.2.44)
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PNPP Chapter 3, Table 3.1.2-2, Pages 3.1-47 thru 3.1-55, as supplemented by Supplement 2, Attachment 9, Vistra Letter L-24-020 (pdf pages 110 -118) is revised as follows: (see following pages)
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 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 B
8 Filter housing Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
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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
10 12 Flexible hose Pressure boundary Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 11 Flexible hose Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
Cracking External Surfaces Monitoring of Mechanical Components IV.E.RP-04 3.1.1-107 H,
111 12 13 Flexible hose Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
13 14 Flexible hose Pressure boundary Stainless steel Air - indoor, uncontrolled (Int)
None None VII.J.AP-123 3.3.1-120 A
14 15 Orifice Flow restriction, Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
15 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
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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 16 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
17 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
18 19 Orifice Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
19 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
20 21 Piping Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
21 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
22 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
23 24 Piping Leakage boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
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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 24 25 Piping Leakage boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
25 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
26 27 Piping Pressure boundary Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 27 28 Piping Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
28 29 Piping Pressure boundary Stainless steel Air - indoor, uncontrolled (Int)
None None VII.J.AP-123 3.3.1-120 A
29 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
30 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
31 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
32 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
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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 33 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
34 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
35 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
36 37 Piping Pressure boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
37 38 Piping Pressure boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
38 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
39 40 Piping Structural integrity Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 40 41 Piping Structural integrity Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
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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 41 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
42 43 Piping Structural integrity Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
43 44 Piping Structural integrity Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
44 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
45 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
46 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
47 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
48 49 Strainer body Leakage boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
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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 49 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
50 51 Valve body Leakage boundary Copper alloy >15%
Zn Air - indoor, uncontrolled (Ext)
None None V.F.EP-10 3.2.1-57 A
51 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
52 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
53 54 Valve body Leakage boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
54 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
55 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
56 57 Valve body Leakage boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
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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 57 58 Valve body Leakage boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
58 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
59 60 Valve body Pressure boundary Aluminum Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-134 3.3.1-113 E, 105 60 61 Valve body Pressure boundary Aluminum Air - indoor, uncontrolled (Ext)
None None VII.J.AP-135 3.3.1-113 A
61 62 Valve body Pressure boundary Stainless steel Air - dry (Int)
None Compressed Air Monitoring VII.J.AP-20 3.3.1-120 E, 105 62 63 Valve body Pressure boundary Stainless steel Air - indoor, uncontrolled (Ext)
None None IV.E.RP-04 3.1.1-107 A
63 64 Valve body Pressure boundary Stainless steel Air - indoor, uncontrolled (Int)
None None VII.J.AP-123 3.3.1-120 A
64 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
65 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
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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 66 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
67 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
68 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
69 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
70 71 Valve body Pressure boundary Steel Treated water (Int)
Cumulative fatigue damage TLAA IV.C1.R-220 3.1.1-6 A
71 72 Valve body Pressure boundary Steel Treated water (Int)
Loss of material Flow-Accelerated Corrosion IV.C1.R-23 3.1.1-60 A
72 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 8 L-25-013 Attachment 2 Page 1 of 3 LRA Section: Table 3.1.1 Notes for Table 3.1.2-1 through Table 3.1.2-6 LRA Page Number(s): Pages 3.1-110 and 3.1-111
References:
NCSG RAI-10276-R1, Vistra Letter L-24-208, Attachment 2 Description of Change: Reference Attachment 1 for the bases and description of the LRA update to plant specific Note 111.
PNPP Chapter 3, Notes for Table 3.1.2-1 through Table 3.1.2-6, Pages 3.1-109 and 3.1-110, as supplemented by the first Annual Update, Attachment 7, Vistra Letter L-24-110 (pdf pages 71 and 72) is revised as follows: (see following pages)
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 2 Page 2 of 3 Notes for Table 3.1.2-1 through Table 3.1.2-6 Standard Notes A. Consistent with NUREG-1801 item for component, material, environment and aging effect. AMP is consistent with NUREG-1801 AMP.
B. Consistent with NUREG-1801 item for component, material, environment and aging effect. AMP takes some exceptions to NUREG-1801 AMP.
C. Component is different, but consistent with NUREG-1801 item for material, environment and aging effect. AMP is consistent with NUREG-1801 AMP.
D. Component is different, but consistent with NUREG-1801 item for material, environment and aging effect. AMP takes some exceptions to NUREG-1801 AMP.
E. Consistent with NUREG-1801 item for material, environment and aging effect, but a different aging management program is credited or NUREG-1801 identifies a plant-specific aging management program.
F. Material not in NUREG-1801 for this component.
G. Environment not in NUREG-1801 for this component and material.
H. Aging effect not in NUREG-1801 for this component, material, and environment combination.
I. Aging effect in NUREG-1801 for this component, material and environment combination is not applicable.
J. Neither the component nor the material and environment combination are evaluated in NUREG-1801.
Plant-Specific Notes 101 Main steam line venturi material is A-351 grade CF8 (low molybdenum), with ferrite <20%. Based on the criteria set forth in the May 19, 2000, NRC (Grimes) letter, this material is not susceptible to thermal embrittlement.
102 High component surface temperature precludes moisture accumulation that could result in corrosion.
103 The internal environment of these components is Fyrquel', a phosphate ester used as a hydraulic fluid. For the purposes of aging evaluation, this environment is compared to the GALL environment of "Lubricating oil."
104 See LRA section 4.6.2, Main Steam Line Flow Restrictors Erosion Analysis.
105 The Compressed Air Monitoring program provides assurance that the quality of the "Air - dry" environment supports the conclusion that no aging effects are expected.
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 2 Page 3 of 3 106 Conservatively assumed as high strength steel. See Reactor Head Closure Stud Bolting Program.
107 This component type represents high strength, low alloy steel flange bolts servicing the Reactor Head Spray flange (N8). High-strength steel and steel bolting in the environments of air with reactor coolant leakage, Air, and System temperature up to 288°C (550°F) are considered equivalent for the aging effects of Loss of Material, Loss of Preload, and Cumulative Fatigue Damage.
108 This component type represents the CRD housings that consists of stainless steel and nickel components with the nickel pipe welded to the reactor vessel head.
109 Loss of material for the jet pump restrainer bracket and cast austenitic stainless steel jet pump wedges surface is aligned to GALL line item to address wear.
110 These components are the stainless steel boundary valves in the Reactor Coolant Pressure Boundary System exposed to sodium pentaborate solution.
111 Per Table 4-1, EPRI Report 1010639, Non-Class 1 Mechanical Implementation Guideline and Mechanical Tool.
Revision 4, Appendix D, Nickel-Base Alloy aging effects are treated same as stainless steel in a dried air environment. The Compressed Air Monitoring program provides assurance that the quality of the "Air - dry" environment supports the conclusion that no aging effects are expected.
111 NUREG-1801 provides no aging effects for this component type, material, and environment combination.
PNPP assigns Cracking due to various causes such as installation-initiated cracking exacerbated by chloride induced SCC, chloride induced SCC, and / or cycle fatigue is assigned to this row. Two flexible hoses are associated with each Safety Relief Valve (SRV). This row represents the external surface of the flexible hoses supplying compressed air (Air - dry (Int) environment) to the operator of each Safety Relief Valve (SRV) in the main steam lines. These flexible hoses will be replaced initially at a 3-cycle frequency and will extend the service lives into the PEO.
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 3 Page 1 of 3 LRA Section: Table 3.1.1, Item 3.1.1-107 LRA Page Number(s): Page 3.1-144
References:
NCSG RAI-10276-R1, Vistra Letter L-24-208, Attachment 2 Description of Change: Reference Attachment 1 for the bases and description of the LRA update to Table 3.1.1, Item 3.1.1-107.
PNPP Chapter 3, Table 3.1.1, Item 3.1.1-107, Page 3.1-144 is revised as follows: (see following pages)
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Table 3.1.1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Components in Chapter IV of NUREG-1801 Item Number Component Aging Effect/
Mechanism Aging Management Programs Further Evaluation Recommended Discussion 3.1.1-106 Nickel alloy piping, piping components and piping element exposed to air -
- indoor, uncontrolled, or air with borated water leakage None None NA - No AEM or AMP Consistent with NUREG-1801.
3.1.1-107 Stainless steel piping, piping components and piping element exposed to gas, concrete, air with borated water leakage, air -
- indoors, uncontrolled None None NA - No AEM or AMP Consistent with NUREG-1801, with the following clarification and a different program. The External Surfaces Monitoring of Mechanical Components program will manage cracking due to various mechanisms of stainless steel flexible hoses supplying compressed air to safety relief valves in the Nuclear Boiler system.
See Appendix B, Section B.2.18. In addition to the Reactor Vessel, Internals, and Reactor Coolant systems; stainless steel commodities in concrete in the Bulk Civil Commodities, Containment Structure, and Turbine Buildings and
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Table 3.1.1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Components in Chapter IV of NUREG-1801 Item Number Component Aging Effect/
Mechanism Aging Management Programs Further Evaluation Recommended Discussion Associated Structures, Process Facilities, and Yard Structures have been aligned with this item.
3.1.1-108 There is no Item Number 3.1.1-108 in NUREG-1800 Rev. 2 or subsequent ISGs.
3.1.1-109 There is no Item Number 3.1.1-109 in NUREG-1800 Rev. 2 or subsequent ISGs.
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 4 Page 1 of 2 LRA Section: A.1.18 LRA Page Number(s): Pages A-19 thru A-20
References:
NCSG RAI-10276-R1, Vistra Letter L-24-208, Attachment 2 Description of Changes: Reference Attachment 1 for the bases and description of the LRA update to Appendix A, Section A.1.18.
PNPP Section A.1.18, Pages A-19 thru A-20, as modified by Attachment 41, Vistra Letter L-24-020 (pdf pages 322 thru 323) and Attachment 12, Vistra Letter L-24-256 (pdf pages 194 thru 195) is revised as follows:
A.1.18 EXTERNAL SURFACES MONITORING OF MECHANICAL COMPONENTS 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 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/
Mineral fiber) materials and cracking and loss of material in aluminum 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.
PNPP does not have an air environment containing halides. Stainless steel flexible hoses supplying compressed air to the main steam line safety relief valves located in the Drywell have experienced cracking on the external surface. These surfaces, typically exposed to air, have been periodically exposed to a local source of chloride contaminates from a leak detection solution. The external surface of the pressure boundary for these flexible hoses is physically inaccessible. It is enclosed in a stainless steel mesh integral to its structural integrity. Consequently, these components will be managed by periodically replacing them at a frequency that
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 4 Page 2 of 2 provides reasonable assurance the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.
The program is also credited with managing loss of material from internal surfaces of metallic components and with loss of material, cracking, and change in material properties from the internal surfaces of elastomers, for cases in which material and environment combinations are the same for internal and external surfaces such that external surface condition is representative of internal surface condition.
The inspections of external surfaces will be capable of detecting age-related degradation.
Surfaces that are inaccessible or not readily visible during either normal plant operations or refueling outages will be inspected opportunistically and at such intervals that will ensure the components intended function is maintained during the period of extended operation. Surfaces that are accessible will be inspected at an interval not to exceed one operating cycle. Inspections will be performed by personnel qualified through plant-specific programs. Deficiencies will be documented and evaluated under the Corrective Action Program.
Outdoor insulated components, and indoor insulated components exposed to condensation (because the in-scope component is operated below the dew point), will have portions of the insulation inspected or removed, during each 10-year interval of the period of extended operation, to determine whether the exterior surface of the component is degrading or has the potential to degrade.
The program will be implemented no later than six months prior to the period of extended operation. Visual inspection of external surfaces will be conducted once every refueling cycle. A sample of insulated piping will be inspected every ten years during the period of extended operation. Inspections will commence during the period of extended operation.
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 5 Page 1 of 4 LRA Section: B.2.18 LRA Page Number(s): Pages B-63 thru B-64
References:
NCSG RAI-10276-R1, Vistra Letter L-24-208, Attachment 2 Description of Changes: See Attachment 1 for the bases for the description of change to PNPP LRA Appendix B, Section B.2.18.
PNPP Section B.2.18, Pages B-63 thru B-64, as modified in accordance with the response to NCSG RAI 10337-R1, Attachment 15, Vistra Letter L-24-209 which resulted in Supplement 7, Attachment 17, Vistra Letter L-24-256 (pdf pages 214 thru 216) is revised as follows:
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 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.
PNPP does not have an air environment containing halides. Stainless steel flexible hoses supplying compressed air to the main steam line safety relief valves located in the Drywell have experienced cracking on the external surface. These surfaces, typically exposed to air, have been periodically exposed to a local source of chloride contaminates from a leak detection solution. The external surface of the pressure boundary for these flexible hoses is physically inaccessible. It is enclosed in a stainless steel mesh integral to its structural integrity. Consequently, these components will be managed by periodically replacing them at a frequency that
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 5 Page 2 of 4
provides reasonable assurance the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.
The program is also credited with managing loss of material from internal surfaces of metallic components and with loss of material, cracking, and change in material properties from the internal surfaces of elastomers, for cases in which material and environment combinations are the same for internal and external surfaces such that external surface condition is representative of internal surface condition.
The inspections of external surfaces will be capable of detecting age-related degradation.
Surfaces that are inaccessible or not readily visible during either normal plant operations or refueling outages will be inspected opportunistically and at such intervals that will ensure the components intended function is maintained during the period of extended operation. Surfaces that are accessible will be inspected at an interval not to exceed one operating cycle. Inspections will be performed by personnel qualified through plant-specific programs. Deficiencies will be documented and evaluated under the Corrective Action Program.
Outdoor insulated components, and indoor insulated components exposed to condensation, will have portions of the insulation inspected or removed, during each 10-year interval of the period of extended operation, to determine whether the exterior surface of the component is degrading or has the potential to degrade.
The program will be implemented no later than six months prior to the period of extended operation. Visual inspection of external surfaces will be conducted once every operating cycle. A sample of insulated piping will be inspected every ten years during the period of extended operation. Inspections will commence during the period of extended operation.
NUREG-1801 Consistency The External Surfaces Monitoring of Mechanical Components Program is a new program for PNPP that will be consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.M36, External Surfaces Monitoring of Mechanical Components as revised by LR-ISG-2012-02.
Exceptions to NUREG-1801 None Enhancements None Operating Experience The following operating experience examples provide objective evidence that the External Surfaces Monitoring of Mechanical Components Program will be effective in ensuring that component intended functions are maintained consistent with the current licensing basis during the period of extended operation.
Perry Nuclear Power Plant LRA Supplement 8 L-25-013 Attachment 5 Page 3 of 4
A review of plant-specific operating experience since 2013, through a search of plant Corrective Action Program documents found 138 corrective actions potentially related to XI.M36 or identifying degraded or potentially degraded insulation. A review of the operating experience found:
In September 2016, a Condition Report (CR) documented damaged or missing insulation in the off gas vault refrigeration system.
Numerous instances of component leaks throughout the plant that were obscured by overlying insulation (e.g., in July 2017 a CR documented a leak in the reactor core isolation cooling system from under the insulation). Such leaks could lead to insulation degradation.
In February 2020, a CR documented that a visual inspection found evidence of corrosion of piping or components under insulation in the turbine building chilled water system.
In February 2020, a CR documented corroded piping or components under insulation were found during repairs in the turbine building chilled water system.
All issues were corrected. The operating experience shows that routine walkdowns and inspections do identify degradation of outdoor components and indoor insulation and leaks under insulation that could identify cracks, seal degradation or failed components.
Routine maintenance has shown that rust, flaking and other piping or component degradation does occur under insulation. No evidence of failed or leaking joints in jackets have been identified. Implementation of the External Surfaces Monitoring of Mechanical Components Program will assure system inspections and walkdowns inspect all outdoor accessible components and all indoor accessible insulation each operating cycle. Removal of a sampling of insulation to inspect piping and components will assure degradation from condensation occurring under the insulation will be detected.
In 2021, condition reports document leaking stainless steel flexible air supply hoses for safety relief valves 1B21F0041B and 1B21F0051B. The condition reports cite the suspected failure causes as cyclic fatigue or cracking due to stress corrosion cracking. Both condition reports note that the causes of leaks are not yet known and that a failure analysis is needed to definitively determine the cause. In lieu of a failure analysis, PNPP developed an equivalent replacement hose change document with an improved material resistant to stress corrosion cracking. However, this replacement modification was not implemented. Although failure analyses were ultimately not performed on the flex hoses for the leaks identified in 2021, in 2023 new condition reports identified additional leaking stainless steel flexible air hoses for safety relief valves 1B21F0041B (replaced in 2021) and 1B21F0047B (original equipment). PNPPs operating experience review performed for these condition reports identified that previous leaks in the flexible air hoses for the same system had occurred in 2021 (SRV-0041B and -0051B), in 2017 (SRV-0041B), and in 2011 (SRV-0051B). An outside vendor performed a failure analysis for the leaking flex hoses from 2023, and their failure analysis reports determined the cause to be outside diameter chloride induced stress corrosion cracking. In 2024 an additional condition report was initiated to determine the source of the chloride contamination of the flex hoses, and to prescribe
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corrective actions to ensure component reliability. The source of the chloride contamination was attributed to accumulated residue from leak detector solution in the corrugations of the flex hoses under the stainless steel braided sheathing. The planned corrective actions included a design change to the hose design to reduce the potential for excessive installation stresses, enhanced procedural guidance for hose installation, revised procedural guidance for leakage testing, and the periodic replacement of the hoses. The initial replacement frequency will be every 3 operating cycles.
Conclusion The external surfaces monitoring of mechanical components program will provide reasonable assurance that aging effects will be managed such applicable components will continue to perform their intended functions consistent with the current licensing basis during the period of extended operation.