NL-25-0152, Propoosed Inservice Alternative FNP-ISI-ALT-05-13 for Reactor Vessel Bottom Head Inspection
| ML25118A329 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/28/2025 |
| From: | Chamberlain A Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-25-0152 | |
| Download: ML25118A329 (1) | |
Text
~ Southern Nuclear April 28, 2025 Docket No.:
50-364 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, D. C. 20555-0001 Regulatory Affairs Joseph M. Farley Nuclear Plant - Unit 2 3535 Colonnade Parkway Binningham, AL 35243 205 992 5000 NL-25-0152 Proposed I nservice Inspection Alternative FN P-I SI-AL T-05-13 for Reactor Vessel Bottom Head Inspection Ladies and Gentlemen:
In accordance with 10 CFR 50.55a, "Codes and standards," Southern Nuclear Operating Company (SNC) requests NRC approval of a proposed alternative associated with the lnservice Inspection (ISi) Program for Farley Nuclear Plant (Farley). Pursuant to 10 CFR 50.55a(z)(2), an alternative is being requested on the basis that a hardship without a compensating increase in quality and safety exists for the supplemental examination of the Reactor Pressure Vessel (RPV) bottom-mounted instrument penetrations and instrument connections as required by ASME Section XI IWB-3142.2.
The proposed alternative is further described in the attached enclosure. Currently, the scheduled start of work on the A Train Residual Heat Removal Seal Cooler is delayed as is the proposed work on the Polar Crane to make health improvements until acceptance of this alternative. In addition, due to the complexity of the high-risk activities and specialty skills and qualifications required to perform the lower internals removal, mobilizing of resources would need to start within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or it would further challenge the outage critical path. Therefore, SNC requests approval of this alternative within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of submittal.
This letter contains no NRC commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.
Respectfully submitted,
~~
Amy Chamberlain Licensing Manager Southern Nuclear Operating Company ACC/was/cgb
U.S. Nuclear Regulatory Commission NL-25-0152 Page 2
Enclosure:
FNP-ISI-AL T-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2) cc:
Regional Administrator, Region II NRR Project Manager - Farley 1 & 2 Senior Resident Inspector - Farley 1 & 2 RType: CFA04.054
Joseph M. Farley Nuclear Plant - Unit 2 Proposed lnservice Inspection Alternative FNP-ISI-AL T-05-13 for Reactor Vessel Bottom Head Inspection Enclosure FNP-ISI-AL T-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2)
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2)
Plant Site - Unit:
Interval-Interval Dates:
Requested Date for Approval:
ASME Code Components Affected:
Applicable Code Edition and Addenda:
Applicable Code Requirements:
Reason for Request:
Farley Nuclear Plant (FNP) Unit 2 5th ISi Interval extending from December 1, 2017, through November 30, 2027.
Approval is requested by April, 30, 2025, to support the Spring 2025 Outage (2R30) at FNP-2.
ASME Code Case N-722-1, ASME Items B15.80 and B15.100 for RPV bottom-mounted instrument penetrations and instrument connections ASME Section XI, 2007 Edition through the 2008 Addenda.
10 CFR 50.55a(g)(6)(ii)(E) mandates use of ASME Code Case N-722-1.
Code Case N-722-1 requires a visual examination of ASME Items B15.80 and B 15.100 for RPV bottom-mounted instrument penetrations and instrument connections. Inspections acceptance standards are in accordance with ASME Section XI IWB-3522 which provides additional acceptance standards for relevant conditions in accordance with IWB-3142.
Pursuant to 10 CFR 50.55a(z)(2), an alternative is being requested by Southern Nuclear Operating Company (SNC) for Farley Nuclear Plant (FNP) Unit 2 on the basis that a hardship and unusual difficulty exists in the supplemental examinations of the RPV bottom-mounted instrument penetrations and instrument connections as required by ASME Section XI IWB-3142.2. While performing the visual examinations required by ASME Code Case N-722-1, boric acid was present in the annulus areas at penetrations 1, 2, 5, 8, 12, 13, 17, 19, 20, 22, 23, 26, 28, 29, 36, 37, 39, 42, 43, and 46 for ASME Item B15.80. Additionally, boric acid was present on the alloy 600 tube to safe end welds at penetration locations 1-6, 8, 9, 10, 13, 17, 19, 20, 22, 26, 28, 29, 36, 37, 43, and 46 for ASME Item number B15.100 visual examinations. ASME Code Case N-722-1 requires acceptance of this condition in accordance with IWB-3522. This is a relevant condition per IWB-3522; therefore, IWB-3142 and IWA-5250 E-1
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2) are applicable. Per IWB-3142, Acceptance can be performed by the following methods:
IWB-3142.1 Acceptance by Visual Examination (a) A component whose visual examination confirms the absence of the relevant conditions described in the standards of Table IWB-3410-1 shall be acceptable for continued service.
(b) A component whose visual examination detects the relevant conditions described in the standards of Table IWB-3410-1 shall be unacceptable for continued service, unless such components meet the requirements of IWB-3142.2, IWB-3142.3, or IWB-3142.4.
IWB-3142.2 Acceptance by Supplemental Examination.
A component containing relevant conditions shall be acceptable for continued service if the results of supplemental examinations (IWB-3200) meet the requirements ofIWB-3130.
IWB-3142.3 Acceptance by Corrective Measures or Repair/Replacement Activity. A component containing relevant conditions is acceptable for continued service if the relevant conditions are corrected by a repair / replacement activity or by corrective measures to the extent necessary to meet the acceptance standards of Table IWB-3410-1.
IWB-3142.4 Acceptance by Analytical Evaluation. A component containing relevant conditions is acceptable for continued service if an analytical evaluation demonstrates the component's acceptability. The evaluation analysis and evaluation acceptance criteria shall be specified by the Owner. A component accepted for continued service based on analytical evaluation shall be subsequently examined in accordance with IWB-2420(b) and (c).
IWB-3142.1 is not possible due to the presence of boric acid in the annulus region and on the alloy 600 tube to safe end welds at the above noted locations. IWB-3142.3, Corrective measures or repair/replacement activities are not applicable because there has been no physical flaw identified in any of the penetrations. IWB-3142.4, Acceptance by Analytical Evaluation, cannot be performed because there is no definitive method to determine if some boric acid was present due to inservice conditions prior to cavity seal leakage present following refueling cavity flood up to support refueling operations. Performance of IWB-3142.2 was considered to perform volumetric examinations of the subject penetrations; however, performance of supplemental examinations would result in a hardship without a compensating increase in the level of quality and safety.
To set proper conditions for the supplemental examinations it is necessary to disassemble the reactor vessel, offload the core, remove the E-2
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2)
Proposed Alternative and Basis for Use:
upper and lower internals (core barrel) which results in high-risk tasks and substantial unplanned dose. High risk tasks include emergently lifting the reactor head, removal of the upper internals, fuel offload, removal of the lower internals, and returning all these components back into the vessel for start-up. All these tasks carry the risks of damaging components during the emergent evolution. Furthermore, the lower internals must be removed to perform supplemental exams on the reactor vessel bottom-mounted instrument penetrations and instrument connections from inside the reactor vessel. The supplemental examinations must be performed internally due to the orientation of the J-groove weld. The lower internals lift will disrupt 2R30 outage projects during the lower internals lift due to the high dose of the component and potential exposure to employees on the 155 ft elevation (refueling floor) of containment. The estimated total dose for all the reactor vessel disassembly, the supplemental exams from the inner diameter, and the reactor vessel reassembly is 6.483 Person Rem if performed emergently during 2R30, this would be a 33% increase to total outage dose. If outer diameter (OD) volumetric examinations were performed on the impacted ASME Item number B15.100 exams alone due to their access from the OD, estimated dose for performance of this exam is 3.066 Person Rem. Therefore, the significant increase to personnel dose and safety risk to the plant that could result from performing the supplemental examinations required by ASME Section XI IWB-3142.2 represents a hardship.
Background
During 2R30 shutdown, a pre flood up visual inspection was completed of the vessel lower head with a qualified VT-2L examiner. The inspection was performed around the full circumference of the reactor vessel lower head, observed from two locations 180 degrees from each other. The inspection was performed similarly to the inspection performed during the Class 1 leakage test performed at the end of each refueling outage with the reactor pressure vessel pressurized to nominal operating pressure.
This visual inspection did not look at individual penetrations and did not meet the individual penetration and tube to safe end welds visual examinations per the requirements of ASME Code Case N-722-1; however, this examination did confirm that there was no presence of boric acid from the high level overview of the bottom of the vessel.
Additionally, there was no evidence of leakage coming from above the bottom of the reactor vessel.
During refueling, the refueling cavity is flooded to support refueling activities. The cavity seal is designed to provide a leak tight barrier while the cavity is flooded to prevent refueling water from coming in contact with the reactor vessel exterior. Following removal of the cavity seal and subsequent refueling activities, a leakage path was identified as the E-3
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2) cause of boric acid impacting the bottom vessel inspection areas. During the previous refueling outage, 2R29, cavity seal leakage was identified while the cavity was flooded resulting in a similar condition. Corrective actions taken after 2R29 cavity seal leakage were unsuccessful in preventing leakage during 2R30. There was no evidence of Reactor Coolant System (RCS) pressure boundary leakage; therefore, cleaning was conducted, and a visual examination was performed of the post cleaning bottom of vessel. This inspection confirmed there was no loss of material in the carbon steel annulus region of the affected penetrations.
Proposed Alternative SNC has completed cleaning of the bottom of the reactor vessel and all instrumentation connections above the insulation package to remove all boric acid residue. In addition, SNC performed a visual baseline examination of the cleaned areas to confirm no loss of material in the carbon steel annulus region during the current refueling outage.
Furthermore, SNC proposes to perform a visual examination of the bottom of the vessel above the insulation during the Class 1 system leakage test (at nominal operating pressure) to confirm that there is no active reactor vessel pressure boundary leakage occurring in this vicinity during this refueling outage. This will examine the accessible external exposed surfaces of pressure retaining components for evidence of leakage by a VT-2L qualified individual. It is important to note, that the ASME Code Case N-722-1 Item B15.100 welds are easily accessible for visual examination from above the insulation package, without complete removal of insulation. Additionally, SNC proposes to perform visual examinations of the bottom vessel (ASME Item 15.80 and 15.100) prior to reactor cavity flood up during 2R31 and will continue each refueling outage until cavity seal leakage has been completely corrected.
Basis for Use Visual Inspections Pre-flood up Visual inspection of the reactor vessel bottom head from two locations 180 degrees apart were performed, prior to the refueling cavity flood up during 2R30, and did not show any signs of leakage from above or any gross evidence of boric acid. The previous two ASME Code Case N-722-1 examinations were completed satisfactorily during 2R26 and 2R28, and there was no evidence of leakage during the previous Class 1 leakage test at the end of the 2R29 refueling outage.
Strategies for Inspection MRP-206, Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants, discusses the value of performing supplemental visual examinations in conjunction with the bare E-4
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2) metal visual of the bottom mounted instrumentation. Farley currently performs supplemental examinations each refueling outage above the insulation with the Reactor depressurized.
Correlation of Cavity Seal Leakage with Location of Leakage Noted on Bottom Head Following refueling cavity draining down during 2R30, non-destructive examination (NOE) personnel performed a visual examination on 4/24/25 in accordance with Code Case N-722-1 and determined that 20 nozzles had boric acid at the annulus area and 21 had boric acid at the alloy 600 tube to safe end welds (dissimilar metal welds). NOE personnel also noted boric acid streaks at approximately 155 degrees and then again from 190 degrees to 270 degrees around the reactor vessel with pooling noted in the middle on top of the insulation package.
Inspection of the segmented cavity seal ring and flange area during 2R30 on 4/25/25 identified water and rust colored staining under the sealing surface from 75 degrees to 225 degrees which aligns with where NOE personnel identified boric acid staining along the outside of the vessel wall on the bottom head, at 155 degrees and between 190 degrees and 270 degrees. Defects were also noted near 200 degrees on the outside of the reactor flange. The cavity seal ring inspection also noted rust colored residue at approximately 45 degrees, but there was no indication of water trailing down the exterior of the vessel in this area which again aligns with what NOE observed on the bottom of the vessel.
Due to the cavity seal ring leakage running down the exterior of the reactor vessel, twenty-five bottom mounted instrumentation (BMI) penetrations had boric acid staining and/or residue in the annulus region, on the dissimilar metal welds, or a combination thereof. All affected penetrations were rejectable indications during the N-722-1 examination in 2R30 and could not be evaluated due to the amount of accumulation seen in the annulus region or trailing down the stalk and over the dissimilar metal welds. There were five BMls that only noted rejectable indications on the dissimilar metal welds and all five were in the center of the vessel. This aligns with the pooling of water at the center of the Reactor vessel on top of the insulation package. This is indicative of water running down the side and onto the middle of the reactor vessel, falling onto the insulation package, and splashing back up onto the stalk where the dissimilar metal weld is located.
During the previous refueling outage, 2R29, cavity seal leakage was identified while the cavity was flooded resulting in a similar condition as seen this refueling outage. Corrective actions included benchmarking VC Summer and implementing the following: using a 0.015" feeler gauge between the seal and floor at each joint, using a mirror to inspect the backside of the seal, validating the bulging of the seal appears even 360 degrees around, and cam alignment and actuation was rotated 90 E-5
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2) degrees (previously cam alignment and actuation was perpendicular to the seal, this was changed to parallel to the seal in 2R30). Just In Time Training (JITT) was led by two industry Subject Matter Experts (SME) prior to 2R30. The two SM Es were also present in the cavity during installation. A Quality Control (QC) inspection of the sealing surface was also performed prior to installation of the cavity seal ring. These actions did not eliminate leakage in 2R30.
Characteristics of Limiting Locations vs OE Plants with Alloy 600 Leakage Limiting locations from the precleaned bottom head were compared to pictures of the OE plants which have experienced reactor coolant system (RCS) pressure boundary leakage from the bottom of the Reactor vessel.
These Farley Unit 2 most limiting penetrations did contain boron in the anulus region. The boron was flat with no appreciable volume. The OE plants contain boric acid that was white with volume, popcorn like boron.
Based on this comparison, the characteristics of the annulus region at Farley are not consistent with what would be expected for RCS pressure boundary leakage based on the comparison with the OE plants.
Operational Rad Monitoring A review of radiation monitoring trends for R11, Containment Air Particulate Monitor, and R12, Containment Gas Monitor, over the past 18 months has shown no significant data to suggest RCS pressure boundary leakage. There were only two spikes throughout the entire cycle. One spike occurred during a filter change out which can be attributed to less filtered air going through the detector and another occurred when the Reactor Coolant Drain Tank Relief Valve lifted and subsequently reseated. In addition, filter swaps were performed under routine preventative maintenance work orders with no additional swaps required.
The Unit 2 R11 particulate and R12 noble gas containment atmosphere radiation monitoring systems (RMSs) are designed and operated to meet the requirements of RCS leak detection equipment. The Farley R11/R12 radiation monitoring systems meet the requirements of 10 CFR 50 Appendix A GDC 30, Regulatory Guide 1.45 Revision 0, which provides guidance on acceptable methods for leak detection instrumentation selection, and Farley Technical Specifications 3.4.15, RCS Leakage Detection Instrumentation.
The original R11/R12 radiation monitors were replaced as part of RMS (radiation monitoring system) modernization efforts. Given changes in technology and detector design as compared to the original system, the current R 11 /R 12 systems have significantly wider detection ranges. One example is the change of the R11 detection assembly from a detector sensitive to gamma radiation to a detector sensitive primarily to beta radiation. With reference to the system descriptions in Chapter 5 of older revisions of the Farley FSAR, the previous measurement ranges for the E-6
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2)
R 11 and R 12 RMSs were 10-9 µCi/cc to 1 o-6 µCi/cc and 1 o-6 µCi/cc to 10-3
µCi/cc, respectively. With the upgraded R11 and R12 RMSs the measurement ranges have widened to 10-12 µCi/cc to 10-6 µCi/cc and 1 o-12 µCi/cc to 10-3 µCi/cc respectively (FSAR Section 5.2. 7.1.1 ). This is an increase in the level of detection by a factor of 1,000 for R11 and 1,000,000 for R12. In other words, R11 now detects radiation levels 1,000 times lower than before. The R12 system's sensitivity increased even more, detecting levels 1,000,000 times lower. These improvements mean the systems can detect much smaller amounts of radiation. The expanded detection range gives us confidence that the systems can identify even small leaks in the containment atmosphere.
During the recent operating cycle, there were no unexplained high readings, indicating no undetected leaks. This provides reasonable assurance of the systems' effectiveness in identifying potential leaks. In summary, the R11 and R12 systems are highly sensitive and dependable for detecting potential leaks, meeting all required standards, and providing assurance of their performance.
RCS Leakage Monitoring A review of unidentified RCS leak rate trends over the past 18 months was performed and the highest leak rate identified was 0.017 gpm on 2/22/25. There were two anomalous days on 11/11/23 of 0.05 gpm and 9/30/24 of 0.066 gpm. No trend or other indication of leakage was identified either day and the unidentified leak rate returned to near zero the following day in both instances.
The Containment sump pumps auto start and stop as needed. Comparing the frequency of pumping from the last 2 cycles, this past cycle does not show an increase in frequency of the pumps running or indicate an RCS leak.
SNC trends RCS leak rate values in accordance with procedures consistent with the guidance of WCAP-16465-NP. These guidelines for leak rate monitoring require a response in the following cases:
Absolute leak rate action levels:
o One seven day rolling average of daily Unidentified RCS leak rates > 0.1 gpm.
o Two consecutive daily Unidentified RCS leak rates >
0.15 gpm.
o One daily Unidentified RCS leak rate > 0.3 gpm.
Deviation from baseline mean action levels:
o Nine consecutive daily Unidentified RCS leak rates >
baseline mean [µ].
o Two of three consecutive daily Unidentified RCS leak rates > [µ + 2cr].
o One daily Unidentified RCS leak rate > [µ +3cr].
Total integrated unidentified leakage action levels.
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FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2) o Short Term (30 days) Total Integrated Unidentified Leakage > 5,000 gallons.
o Long Term (Operating cycle) Total Integrated Unidentified RCS Leakage > 50,000 gallons.
If any of the above action levels are met, station and fleet procedures (FNP-2-SOP-1.12 "RCS Leakage Troubleshooting Guide" and NMP-OS-009 "RCS Unidentified Leakage Assessment") require assessment and response to the increased leakage. The average leak rate for the last 250 days prior to the refueling outage was 0.003 gpm. With an average leak rate this low, deviations from the baseline mean will be readily detectable.
Additionally, if an unidentified RCS leak is greater than 1 gpm or if an identified RCS leak is greater than 1 O gpm, the plant Technical Specification (TS) 3.4.13, RCS Operational Leakage, outlines the timely actions required to maintain safe operability for recovery up to and including a shutdown. In addition to periodic RCS leakage calculations, containment atmosphere particulate radioactivity monitor and at least one containment air cooler condensate level monitor or one containment atmosphere gaseous radioactivity monitor are required to be operable per plant TS 3.4.15, RCS Leakage Detection Instrumentation. The particulate detector required by TS 3.4.15 was previously a gamma detector with a lower measurement range of 1x10-9 µCi/cc. This detection channel was upgraded to a beta detector with a much lower measurement range of 1x10-12 µCi/cc. This is an improvement of the detection capability by a factor of 1000. The previous gamma detector met the requirements to detect a leak of less than 1 gpm in accordance with Regulatory Guide 1.45.
Chemistry Analysis On April 24, 2025, at 11 :45 PM, Chemistry obtained samples from the lower reactor vessel on penetrations 3, 4, 6, 12 and 26 by taking scrapings of boric acid deposits in various locations for analysis. Analysis conducted were gamma isotopic analysis, pH analysis, and boron titrations. The isotopic analysis focused on comparing the ratios of Cobalt isotopes, specifically Co-58 and Co-60, found in the samples. These ratios were compared to those observed during online reactor coolant analysis and after reactor shutdown periods. The analysis was limited to radiocobalt ratios because the samples did not contain Cesium isotopes, Cs-134 and Cs-137. During the previous Unit 2 fuel cycle prior to the Refueling Outage (RFO), the RCS Co-58/Co-60 ratio average was approximately 8.8. The samples collected showed Co-58/Co-60 ratios ranging from 3.12 to 3.91. These ratios are closer to those observed during the current Unit 2 RFO shutdown period, which averaged 4.4 during the Cavity flood-up period of 2R30.
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FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2)
Duration of Proposed Alternative:
Precedents:
Based on the analysis conducted by Farley chemistry personnel, the samples collected from the lower reactor vessel on April 24, 2025, are consistent with refueling pool water from the recent outage. This conclusion is supported by multiple factors, including the absence of short-lived fission product isotopes in the residue samples, the presence of Cr-51, results from boron titration and pH analysis, and the Co-58/Co-60 ratios in the samples, which are similar to those observed in the reactor coolant system during the cavity flood-up period of the 2R30 refueling outage. All analysis techniques were performed in accordance with station procedures and the evaluation methodology outlined in WCAP-15988-NP, Rev. 1.
Cleaning of the Bottom of the RPV Head On 4/25/25, Radiation Protection technicians performed cleaning activities to remove boron from the bottom of the vessel, the bottom mounted instrumentation, including the areas where borated water had leaked, dripped, and splashed. This activity took approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to complete, which was substantially less than projected. The cleaning activities included using warm water to wash off the boron of the associated areas. No scrubbing was required and the post cleaning inspection demonstrated satisfactory cleaning. The ease of cleaning the boron off the associated areas is indicative that the boron did not have time to interact with the surface for a significant duration or in the presence of heat.
Summary The proposed examinations performed prior to cavity flood up, along with investigations and conclusions from analysis, provide reasonable assurance that the boric acid observed is not a result of pressure boundary leakage. In addition, the proposed future examinations to occur during this refueling outage will provide further reasonable assurance of structural integrity until the next refueling outage where re-examination of the RPV bottom-mounted instrument penetrations and instrument connections will occur in accordance with the requirements of ASME Code Case N-722-1.
This alternative will remain in place until the completion of the 2R31 refueling outage, currently scheduled to complete November 15, 2026.
Palo Verde Nuclear Generating Station Unit 1, Docket No. STN 50-528, Relief Request 57 - Request for Alternative to American Society of Mechanical Engineers Code Case N-729-4 for Replacement Reactor E-9
FNP-ISI-ALT-05-13, Version 1.0, Proposed Alternative to ASME Section XI IWB-3142.2 for Acceptance of Relevant Conditions Identified During ASME Code Case N-722-1 Exams in Accordance With 10 CFR 50.55a(z)(2)
References:
Status:
Vessel Closure Head Penetration Nozzle (ML17299B333); Corresponding Safety Evaluation ML18040A331 American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, 2007 Edition through 2008 Addenda ASME Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 MaterialsSection XI, Division 1 MRP-206, Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants, March 2009 WCAP-16465, Rev. 0, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors Awaiting NRC approval.
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