JAFP-21-0075, Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs
| ML21224A123 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Peach Bottom, Nine Mile Point, Limerick, Clinton, Quad Cities, FitzPatrick, LaSalle  |
| Issue date: | 08/12/2021 |
| From: | David Gudger Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-21-0075, NMP1L3422, RS-21-084 | |
| Download: ML21224A123 (13) | |
Text
200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.55a RS-21-084 JAFP-21-0075 NMP1L3422 August 12, 2021 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket No. 50-333 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-410 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265
Subject:
Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs
Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation (WLI) Partial Penetration Nozzle Repairs August 12, 2021 Page 2 In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(2), Exelon Generation Company, LLC (EGC) requests approval of the attached relief request associated with the repair of water level instrumentation (WLI) partial penetration nozzles on the Reactor Pressure Vessel (RPV).
This relief request provides a repair technique for the WLI partial penetration nozzles. EGC would repair these nozzles by installing a weld pad in accordance with ASME Code Case N-638-x, "Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique,Section XI, Division 1", ASME Code Case N-839-x, "Similar and Dissimilar Metal Welding Using Ambient Temperature SMAW Temper Bead Technique,Section XI, Division 1" or similar code case, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147, and installing a new half nozzle on to the weld pad using a partial penetration weld. This repair technique is referred to as a half nozzle repair. (Note: -x refers to the ASME Code Case revision approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147.)
In support of the flaw evaluation and applicable acceptance criteria associated with the repair, ASME Code Paragraph IWB-3420 and Subarticle IWB-3600 require characterization of the flaw in the partial penetration weld or nozzle. Currently, there is not a Performance Demonstration Initiative (PDI) qualified technique to perform volumetric Non-Destructive Examination (NDE) of the partial penetration weld or nozzle in this configuration that can be used to accurately characterize the location, orientation, or size of a flaw. As an alternative to performing the NDE required to characterize the flaw in the partial penetration weld or nozzle, EGC is proposing to analyze a maximum postulated flaw that bounds the range of flaw sizes that could exist in the original J-groove weld and nozzle, as discussed in.
In accordance with 10 CFR 50.55a(z)(2), EGC proposes the following alternative on the basis that performing a Code required repair results in a hardship without a compensating increase in quality and safety. EGC concludes that the proposed alternative provides an acceptable level of quality and safety.
EGC requests approval of the proposed alternative by August 12, 2022.
contains a summary of commitments.
If you have any questions or require additional information, please contact Tom Loomis at 610-765-5510.
Respectfully, David T. Gudger Sr. Manager - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Summary of Commitments
- 2) Relief Request
Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation (WLI) Partial Penetration Nozzle Repairs August 12, 2021 Page 3 cc:
Regional Administrator - NRC Region I Regional Administrator - NRC Region III NRC Senior Resident Inspector - Clinton Power Station NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - James A. FitzPatrick Nuclear Power Plant NRC Senior Resident Inspector - LaSalle County Station NRC Senior Resident Inspector - Limerick Generating Station NRC Senior Resident Inspector - Nine Mile Point Nuclear Station NRC Senior Resident Inspector - Peach Bottom Atomic Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station NRC Project Manager - Clinton Power Station NRC Project Manager - Dresden Nuclear Power Station NRC Project Manager - James A. FitzPatrick Nuclear Power Plant NRC Project Manager - LaSalle County Station NRC Project Manager - Limerick Generating Station NRC Project Manager - Nine Mile Point Nuclear Station NRC Project Manager - Peach Bottom Atomic Power Station NRC Project Manager - Quad Cities Nuclear Power Station W. DeHass, Pennsylvania Bureau of Radiation Protection A. L. Peterson, NYSERDA Illinois Emergency Management Agency - Division of Nuclear Safety
ATTACHMENT 1 Summary of Commitments
Summary of Commitments Page 1 of 1 Summary of Commitments The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described for the NRC's information and are not regulatory commitments.)
COMMITMENT COMMITTED DATE OR "OUTAGE" COMMITMENT TYPE One-Time Action (Yes/No)
Programmatic (Yes/No)
The final one-cycle flaw analytical evaluation, evaluation of repair, and corrosion evaluation will be submitted within 14 days following the end of the refueling outage in which the flaw is identified.
Within 14 days following the end of the refueling outage in which the flaw is identified.
Yes No
ATTACHMENT 2 RELIEF REQUEST
10 CFR 50.55a RELIEF REQUEST Revision 0 (Page 1 of 7)
Proposed Alternative Associated with One-Cycle Reactor Pressure Vessel Instrument Nozzle Repair in Accordance with 10 CFR 50.55a(z)(2)
- 1. ASME CODE COMPONENTS AFFECTED Code Class:
1
Reference:
IWB-2500, Table IWB-2500-1 Exam Category:
B-P Item Number:
B15.10
==
Description:==
RPV WLI Partial Penetration Nozzles
- 2. APPLICABLE CODE EDITION AND ADDENDA PLANT INTERVAL ASME SECTION XI EDITION START END ASME CONSTRUCTION CODE (RPV)
Clinton Power Station, Unit 1 Fourth 2013 Edition July 1, 2020 June 30, 2030 Section III 1971 Edition, through Summer 1973 Addenda Dresden Nuclear Power Station, Units 2 and 3 Fifth 2007 Edition, through 2008 Addenda January 20, 2013 January 19, 2023 Section III 1963 Edition, through Summer 1964 Addenda Dresden Nuclear Power Station, Units 2 and 3 Sixth 2017 Edition January 20, 2023 January 19, 2033 James A. FitzPatrick Nuclear Power Plant Fifth 2007 Edition, through 2008 Addenda August 1, 2017 June 15, 2027 Section III 1967 Edition, through Winter 1969 Addenda LaSalle County Stations, Units 1 and 2 Fourth 2007 Edition, through 2008 Addenda October 1, 2017 September 30, 2027 Section III 1968 Edition, through Summer 1970 Addenda, except Paragraph N-355 Limerick Generating Station, Units 1 and 2 Fourth 2007 Edition, through 2008 Addenda February 1, 2017 January 31, 2027 Section III 1968 Edition, through Summer 1969 Addenda, except that Article 4 of the Winter 1969 Addenda applies Nine Mile Point Nuclear Station, Unit 1 Fifth 2013 Edition August 23, 2019 August 22, 2029 Section I 1962 Edition Nine Mile Point Nuclear Station, Unit 2 Fourth 2013 Edition October 6, 2018 August 22, 2028 Section III 1971 Edition, through Winter 1972 Addenda Peach Bottom Atomic Power Station, Units 2 and 3 Fifth 2013 Edition January 1, 2019 December 31, 2028 Section III 1965 Edition, through Winter 1965 Addenda Quad Cities Nuclear Power Station, Units 1 and 2 Fifth 2007 Edition, through 2008 Addenda April 2, 2013 April 1, 2023 Section III 1965 Edition, through Summer 1965 Addenda Quad Cities Nuclear Power Station, Units 1 and 2 Sixth 2017 Edition April 2, 2023 April 1, 2033
10 CFR 50.55a RELIEF REQUEST Revision 0 (Page 2 of 7)
- 3. APPLICABLE CODE REQUIREMENTS Flaw Removal Paragraph IWA-4412 states "Defect removal shall be accomplished in accordance with the requirements of IWA-4420." (Applicable to 07/08, 2013, 2017 Editions.)
Subparagraph IWA-4611.1(a) states Defects shall be removed in accordance with IWA-4422.1. A defect is considered removed when it has been reduced to an acceptable size.
(Applicable to 07/08, 2013, 2017 Editions.)
Subarticle IWA-5250(a)(3) states Components requiring corrective action shall have repair/replacement activities performed in accordance with IWA-4000 or corrective measures performed where the relevant condition can be corrected without a repair/replacement activity. (Applicable to 07/08, 2013, 2017 Editions, with minor editorial changes to 2017 Edition.)
Subarticle P-89 of the ASME Code Section l 1962 Edition, Subarticle PW-40 of the ASME Code Section l 1965 Edition up to and including the 2019 Edition, Subarticle N-528 of the ASME Code Section Ill 1963 Edition up to and including the Winter 1970 Addenda, or Subarticle NB-4453 of the ASME Code Section Ill 1971 Edition up to and including the 2019 Edition, requires repair of weld defects, including complete and satisfactory removal of defects detected visually, by examinations, or by leakage tests. (Applicable to 07/08, 2013, 2017 Editions.)
Subarticle P-88 of the ASME Code Section l 1962 Edition, Subarticle PW-39 of the ASME Code Section l 1965 Edition up to and including the 2019 Edition, Subarticle N-532 of the ASME Code Section Ill 1963 Edition up to and including the Winter 1970 Addenda, or Subarticle NB-4620 of the ASME Code Section Ill 1971 Edition up to and including the 2019 Edition, contains requirements for postweld heat treatment.
Flaw Evaluation Subparagraph IWB-3142.1(b) states "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." (Applicable to 07/08, 2013, 2017 Editions.)
Subarticle IWA-3300(a) states, in part, "Flaws detected by the preservice and inservice examinations shall be sized... " (Applicable to 07/08, 2013, 2017 Editions.)
Subarticle IWA-3300(b) states, in part, "Flaws shall be characterized in accordance with IWA-3310 through IWA-3390, as applicable... " (Applicable to 07/08, 2013, 2017 Editions.)
Subarticle IWB-3420 states "Each detected flaw or group of flaws shall be characterized by the rules of IWA-3300 to establish the dimensions of the flaws. These dimensions shall be used in conjunction with the acceptance standards of IWB-3500." (Applicable to 07/08, 2013, 2017 Editions.)
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Subparagraph IWB-3522.1 states, in part, "A component whose visual examination (IWA-5240) detects any of the following relevant conditions shall meet IWB-3142 and IWA-5250 prior to continued service... " (Applicable to 07/08, 2013, 2017 Editions.)
Subarticle IWB-3610(b) states, in part, "For purposes of evaluation by analysis, the depth of flaws in clad components shall be defined in accordance with Fig. IWB-3610-1... "
(Applicable to 07/08, 2013, 2017 Editions, with minor editorial changes to 2013 and 2017 Editions.)
ASME Code Case N-749-x, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147, allows for the use of elastic-plastic fracture mechanics (EPFM) methods in lieu of IWB-3610 and IWB-3620 acceptance criteria to evaluate flaws in ferritic steel components operating in the upper shelf temperature range.
- 4. REASON FOR REQUEST This relief request provides a repair technique for flawed or leaking RPV WLI partial penetration nozzles. This repair partially replaces the existing nozzle assembly with a nozzle that is resistant to Intergranular Stress Corrosion Cracking (IGSCC).
EGC is proposing to apply a welded pad on the Outer Diameter (OD) of the RPV using IGSCC resistant nickel alloy filler metal (for example, 52M). The new weld pad will be welded using the machine Gas Tungsten Arc Welding (GTAW) or manual Shielded Metal Arc Welding (SMAW)
Ambient Temperature Temper Bead (ATTB) welding processes. EGC is proposing to attach an IGSCC resistant nozzle to the new weld pad with a partial penetration weld using a non-temper bead manual welding process and IGSCC resistant nickel alloy filler metal.
The original partial penetration J-groove weld and a remnant of the original nozzle will remain in place. A flaw evaluation will demonstrate the acceptability of leaving the original partial penetration J-groove weld and remnant nozzle, with a maximum postulated flaw, in place for one cycle (see "Flaw Analytical Evaluation" below). Paragraphs IWA-4412 and IWA-4611 contain requirements for the removal of, or reduction in size of defects. The defect on the WLI partial penetration nozzles will not be removed; therefore, relief is sought from these requirements.
Subarticles IWB-3400 and IWB-3600 were written with the expectation that volumetric Non-Destructive Examination (NDE) techniques such as Ultrasonic Testing (UT) would be used to determine the flaw size and shape. In support of the flaw evaluation, the Subarticles IWB-3420 and IWB-3610(b) require characterization of the flaw in the leaking nozzle. Although demonstrated, there is not a Performance Demonstration Initiative (PDI) qualified technique to perform NDE of the partial penetration weld or nozzle in this configuration that can be used to accurately characterize the location, orientation, or size of a flaw.
The flaw evaluation methods presented in Subarticle IWB-3610 and Nonmandatory Appendix A of Section XI are based on Linear Elastic Fracture Mechanics (LEFM) methods. ASME Code Case N-749-x, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147, was developed to provide criteria for the use of Elastic-Plastic Fracture Mechanics (EPFM) methods as acceptable alternatives to the LEFM methods currently contained in Subarticle IWB-3610 and Nonmandatory Appendix A, for operating conditions where ferritic vessel materials are operating on the material toughness upper shelf temperature range.
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Since the Construction Code has requirements for postweld heat treatment, EGC is proposing to install a welded pad in accordance with ASME Code Case N-638-x, N-839-x or similar code case, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147.
- 5. PROPOSED ALTERNATIVE AND BASIS FOR USE In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(2), EGC proposes the following alternative to the requirements specified in Section 3 above on the basis that performing a Code required repair results in a hardship without a compensating increase in quality and safety. This relief request provides a repair technique for the instrument nozzles that will be initially demonstrated for one cycle. EGC intends to follow-up with a relief request for the permanent acceptance of the repair prior to the end of the one cycle. A repair in accordance with the ASME Code, which would remove the flaw from the inner portion of the RPV, would require a full core offload to access the repair location, result in significant risk associated with the inclusion of loose parts and foreign material, and result in significant increase in radiological exposure. These areas of concern result in a significant hardship over the proposed modification.
In lieu of the ASME Code compliant repair, the following alternatives are proposed:
As an alternative to flaw removal or reduction in size to meet the applicable acceptance standards per Paragraphs IWA-4412 and IWA-4611, EGC proposes to implement an OD repair of the RPV WLI partial penetration nozzles utilizing an OD weld pad and half nozzle as described in the repair of nozzle penetration section below.
As an alternative to performing the NDE required to characterize the flaw under Subarticles IWB-3420 and IWB-3610(b) in the WLI partial penetration weld or nozzle, EGC proposes analyzing a maximum postulated flaw that bounds the range of flaw sizes that could exist in the original J-groove weld and nozzle.
As an alternative to the requirements of postweld heat treatment required by the Construction Code, EGC is proposing to install a welded pad in accordance ASME Code Case N-638-x, N-839-x or similar code case, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147.
Basis for Use A. Component Information The RPV and IGSCC susceptible WLI partial penetration nozzles at each plant unit are fabricated as followed:
PLANT RPV MATERIAL (clad with stainless steel)
NOZZLE NUMBER (per unit)1 NOZZLE MATERIAL J-GROOVE WELD MATERIAL Clinton Power Station, Unit 1 SA-533 Grade B Class 1 N12A, N12B, N12C, N12D, N13A, N13B, N13C, N13D, N14A, N14B, N14C, N14D SA-336 Class F8 Alloy 82/182 Dresden Nuclear Power Station, Units 2 and 3 SA-302 Grade B, modified by ASME Code Case 1339 N13A, N13B, N16A, N16B SB-166 (Alloy 600)
Alloy 82/182
10 CFR 50.55a RELIEF REQUEST Revision 0 (Page 5 of 7)
PLANT RPV MATERIAL (clad with stainless steel)
NOZZLE NUMBER (per unit)1 NOZZLE MATERIAL J-GROOVE WELD MATERIAL James A. FitzPatrick Nuclear Power Plant, Unit 1
SA-533 Grade B Class 1, modified by ASME Code Case 1339-2 N11A, N11B, N12A, N12B, N16A, N16B SB-166 (Alloy 600)
Alloy 82/182 LaSalle County Stations, Units 1 and 2 SA-533 Grade B Class 1 N12A, N12B, N12C, N12D, N13A, N13B, N14A, N14B, N14C, N14D SB-166 (Alloy 600)
Alloy 82/182 Limerick Generating Station, Units 1 and 2 SA-533 Grade B Class 1 N11A, N11B, N12A, N12B, N12C, N12D, N16A, N16B, N16C, N16D (Unit 1 only)
SB-166 (Alloy 600)
Alloy 82/182 Nine Mile Point Nuclear Station, Unit 1 SA-302 Grade B, modified N13A, N13B, N14A, N14B, N15A, N15B, N16A, N16B, N17A, N17B SB-166 (Alloy 600)
Alloy 82/182 Nine Mile Point Nuclear Station, Unit 2 SA-533 Grade B Class 1 N12A, N12B, N12C, N12D, N13A, N13B, N14A, N14B, N14C, N14D SA-508 Class 1 Alloy 82/182 Peach Bottom Atomic Power Station, Units 2 and 3 SA-302 Grade B, modified by ASME Code Case 1339 N11A, N11B, N12A, N12B, N16A (Unit 3 only),
N16B SB-166 (Alloy 600)
Alloy 82/182 Quad Cities Nuclear Power Station, Units 1 and 2 SA-302 Grade B, modified by ASME Code Case 1339 N11A, N11B (Unit 1 only),
N12A, N12B SB-166 (Alloy 600)
Alloy 82/182 1 Limerick Generation Station, Unit 2 N-16D, Peach Bottom Atomic Power Station, Unit 2 N-16A, and Quad Cities Nuclear Power Station, Unit 2 N-11B have already implemented this repair There are a total of 103 IGSCC susceptible WLI partial penetration nozzles. EGC intends to repair these nozzles, as required, based on the discussion provided in the following sections.
B. Examination of the J-groove Weld A volumetric (UT) examination will also be performed on the existing J-groove weld in accordance with BWRVIP-03 Revision 19 or later. The current RPV OD volumetric examination technique has been demonstrated to only interrogate the partial penetration J-groove welds and the surrounding RPV low alloy steel (LAS) material. This examination technique provides a method for crack detection, length sizing, and depth sizing of flaws that initiate within the partial penetration J-groove weld material and detects planar flaw indications in the LAS material.
C. Flaw Analytical Evaluation A flaw evaluation for one cycle of operation in accordance with Subarticle IWB-3610 (LEFM method), as well as per ASME Code Case N-749-x (EPFM method) will be performed. ASME Code Case N-749-x or similar code case will be used with all applicable conditions stated in the latest revision of Reg. Guide 1.147.
The ASME Section XI flaw evaluation requires a projection of crack growth for the flaw in the J-groove weld or nozzle being abandoned in-place, and potentially into the RPV LAS material. The potential crack propagation is into the LAS by fatigue and stress corrosion cracking. For this fracture mechanics analyses a conservative evaluation will be performed. This evaluation will be used to demonstrate compliance with a combination of
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Subarticle IWB-3610 and ASME Code Case N-749-x or similar code case (as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147), as applicable. This one-cycle flaw evaluation will be submitted to the NRC. Refer to Summary of Commitments section (Attachment 1) for timing of submittal of the flaw evaluation.
D. Repair of the WLI Partial Penetration Nozzle EGC will replace this existing nozzle assembly with a nozzle penetration that is resistant to IGSCC, using ASME Code Case N-638-x, N-839-x, or similar code case, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147, and in accordance with the construction code. A welded pad will be applied to the OD of the RPV using IGSCC resistant nickel alloy filler metal and will be welded using the machine GTAW or manual SMAW ATTB welding processes. The IGSCC resistant nozzle will be attached to the new weld pad with a partial penetration weld using a non-temper bead manual welding process and IGSCC resistant filler metal. The original partial penetration J-groove weld and a remnant of the original nozzle are planned to remain in place.
The new weld pad will be examined as required by ASME Code Case N-638-x, N-839-x or similar code case, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147. These examinations will verify there are no unacceptable indications in the newly installed weld pad or original base metal material.
A design analysis will be performed in accordance with the design requirements of the construction code. The new design will be reconciled to the original construction code and address all applicable loads to ensure all Code requirements are met. A one-cycle evaluation of the repair will be submitted to the NRC. Refer to the Summary of Commitments section (Attachment 1) for timing of the submittal of the evaluation of the repair.
The current accumulated Effective Full Power Years (EFPYs) and fluence values will be calculated to ensure the material in the area of this repair is not expected to have decreased fracture toughness or ductility associated with damage of the LAS in the beltline region; therefore, there will not be a weldability concern for the repair.
E. Corrosion Evaluation A corrosion evaluation will be performed to consider potential material degradation due to the repair of the RPV WLI partial penetration nozzle. The repair will result in the RPV LAS being exposed to the reactor coolant. The corrosion evaluation will address general corrosion, crevice corrosion, and galvanic corrosion of the exposed LAS in the gap. A review will be performed to determine implementation of the On-Line Noble Metal Chemical addition with Hydrogen Water Chemistry to mitigate corrosion for applicable nozzles. The reactor water chemistry of plants listed in Section 2 meet the requirements of the latest revision of BWRVIP-190 (currently under Revision 1), BWRVIP Water Chemistry Guidelines. The corrosion evaluation will be submitted to the NRC. Refer to Summary of Commitments section (Attachment 1) for timing of submittal of the corrosion evaluation.
F. Loose Parts Evaluations Given the original WLI partial penetration nozzle is not intended to be entirely removed, EGC will complete a lost-parts evaluation to assess the potential for nozzle segments to enter the RPV during power operation. Evaluations will be completed to address the potential impact on the fuel and the potential impact on internal RPV components.
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Conclusion Based on the above, in accordance with 10 CFR 50.55a(z)(2), EGC has concluded that compliance with the ASME Code to perform the repair results in a hardship without a compensating increase in quality and safety. The proposed alternative provides an acceptable level of quality and safety as discussed above.
- 6. DURATION OF PROPOSED ALTERNATIVE Relief is requested for the duration of the operating cycle in which the indication is identified.
A separate relief request will be submitted to justify continued use of the nozzle repair beyond the first cycle. Such relief request, which will contain the appropriate analyses and justification for the remainder of the plant operating life, will be submitted prior to the end of the next upcoming operating cycle.
The requested use of the proposed alternative for each plant in Section 2 is for their current 10-year ISI Interval and the future 10-year ISI Interval for Dresden Nuclear Power Station, Units 2 and 3, and the Quad Cities Nuclear Power Station, Units 1 and 2, which are currently being updated using the 2017 Edition of ASME Section XI as endorsed in the latest revision of 10 CFR 50.55a that will be in effect eighteen months prior to the new interval start dates for those plants.
- 7. PRECEDENTS A similar relief request was previously approved via a verbal authorization on April 15, 2012 for Quad Cities Nuclear Power Station, Unit 2 (ML12107A472). The NRC Safety Evaluation was subsequently issued on January 30, 2013 (ML13016A454). A second similar relief request was previously approved via a verbal authorization on May 17, 2017 for Limerick Generating Station, Unit 2 (ML17137A307). The NRC Safety Evaluation was subsequently issued on August 14, 2017 (ML17208A090). A third similar relief request was previously approved via a verbal authorization on November 6, 2020 for Peach Bottom Atomic Power Station, Unit 2 (ML20314A028). The NRC Safety Evaluation was subsequently issued on April 23, 2021 (ML21110A680).
Another similar repair relief request was approved on January 21, 2016 for Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 concerning alternative requirements for the repair of the reactor Vessel Head Penetrations (VHPs) and J-groove Welds (ML16007A185).