NL-22-0756, Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0

Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0

ML22273A159
Person / Time
Site:	Farley
Issue date:	09/30/2022
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-22-0756
Download: ML22273A159 (21)

Text

Regulatory Affairs 3535 Colonnade Parkway Birmingham AL 35243 205 992 5000 September 30, 2022 Docket Nos.: 50-348 NL-22-0756 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0 Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(z)(1), Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (ISI) alternative FNP-ISI-ALT-05-05, Version 1.0. This proposed alternative, shown in the Enclosure, requests to increase the inspection interval for ASME Section XI, Table IWC-2500-1, exam Category C-B, item number C2.21 and C2.22, exams from 10 years to 20 years through for the remainder of the 5th interval as well as the duration of the 6th ISI Interval.

NRC approval is requested by September 30, 2023, to support Farley refueling outage 2R29.

If you have any questions, please contact Amy Chamberlain at 205.992.6361.

Respectfully submitted, Cheryl A. Gayheart Regulatory Affairs Director CAG/dsp/cbg

Enclosure:

Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) cc: Regional Administrator, Region ll NRR Project Manager - Farley Nuclear Plant Senior Resident Inspector - Farley Nuclear Plant RTYPE: CFA04.054

Joseph M. Farley Nuclear Plant Unit 1 and 2 Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0 Enclosure Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: Class 2

Description:

Steam generator (SG) pressure-retaining welds and full penetration welded nozzles (nozzle-to-shell welds and inside radius sections)

Examination Category: C-B (Pressure Retaining Nozzle Welds in Pressure Vessels,Section XI, Division 1)

Item Numbers: C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections Component IDs:

ASME Unit Component ID Component Description Item No.

1 ALA2-3100-8R 16" MAIN FEEDWATER NOZZLE TO SHELL WELD C2.21 1 ALA2-3200-8R 16" MAIN FEEDWATER NOZZLE TO SHELL WELD C2.21 1 ALA2-3300-8R 16" MAIN FEEDWATER NOZZLE TO SHELL WELD C2.21 1 ALA2-3100-IR8R MAIN FEEDWATER NOZZLE INNER RADIUS C2.22 1 ALA2-3200-IR8R MAIN FEEDWATER NOZZLE INNER RADIUS C2.22 1 ALA2-3300-IR8R MAIN FEEDWATER NOZZLE INNER RADIUS C2.22 2 APR2-3100-8R 16" MAIN FEEDWATER NOZZLE TO SHELL WELD C2.21 2 APR2-3200-8R 16" MAIN FEEDWATER NOZZLE TO SHELL WELD C2.21 2 APR2-3300-8R 16" MAIN FEEDWATER NOZZLE TO SHELL WELD C2.21 2 APR2-3100-IR8R MAIN FEEDWATER NOZZLE INNER RADIUS C2.22 2 APR2-3200-IR8R MAIN FEEDWATER NOZZLE INNER RADIUS C2.22 2 APR2-3300-IR8R MAIN FEEDWATER NOZZLE INNER RADIUS C2.22 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The Fifth Inservice Inspection (ISI) Interval Code of record for Farley Units 1 & 2 is the 2007 Edition through 2008 Addenda of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

3.0 APPLICABLE CODE REQUIREMENT:

ASME Section XI IWC-2500(a), Table IWC-2500-1, Examination Category C-B requires examination of the following Item Nos.:

Item No. C2.21 - Volumetric and surface examination of all nozzle welds at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination area and volume are shown in Figures IWC-2500-4(a), (b), or (d).

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Item No. C2.22 - Volumetric examination of all nozzle inside radius sections at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figures IWC-2500-4(a), (b), or (d).)

4.0 REASON FOR REQUEST:

The Electric Power Research Institute (EPRI) performed an assessment [1] of the basis for the ASME Section XI examination requirements specified for Examination Category C-B of ASME Section XI, Division 1 for Steam Generator (SG) Main Steam (MS) and Feedwater (FW) Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. The assessment includes a survey of inspection results from 74 units as well as flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [1] report concluded that the current ASME Code Section XI inspection interval of ten years can be increased significantly with no impact to plant safety. Based on the conclusions of the EPRI report supplemented by plant-specific evaluations contained herein, Southern Nuclear Company (SNC) is requesting an alternate inspection interval for the subject components. The Reference

[1] report was developed consistent with the recommendations provided in EPRIs White Paper on PFM [11].

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

Farley Unit 1 SNC is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWC-2500-1, Examination Category C-B, Item Nos. C2.21 and C2.22.

The proposed alternative is to increase the inspection interval for these Item Nos. for the Unit 1 replacement SGs to 20 years (from the current ASME Code,Section XI 10-year requirement) for the remainder of the fifth 10-year inspection interval and through the sixth 10-year inspection interval, which is currently scheduled to end on 11/30/2037. All exams will be scheduled to occur in the same period as the last examination, but with a two interval inspection periodicity.

Farley Unit 2 SNC is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWC-2500-1, Examination Category C-B, Item Nos. C2.21 and C2.22.

The proposed alternative is to increase the inspection interval for these Item Nos. for the Unit 2 replacement SGs to 20 years (from the current ASME Code,Section XI 10-year requirement) for the remainder of the fifth 10-year inspection interval and through the sixth 10-year inspection interval, which is currently scheduled to end on 11/30/2037. All exams will be scheduled to occur in the same period as the last examination, but with a two interval inspection periodicity.

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Technical Basis EPRI Technical Report 3002014590 provides the technical basis for this Alternative to ASME Section XI, Table IWC-2500-1, Code Items C2.21 and C2.22.

A summary of the key aspects of the technical basis for this request is provided below.

The applicability of the technical basis of Section 9.0 of Reference [1] to Farley Units 1 &

2 is shown in Appendix A.

In 2000, all three Unit 1 SGs were replaced. The new SG welds and components received the required PSI examinations prior to service followed by ISI examinations through the first period of the current fifth inspection interval.

In 2001, all three Unit 2 SGs were replaced. The new SG welds and components received the required PSI examinations prior to service followed by ISI examinations through the first period of the current fifth inspection interval.

Applicability of the Degradation Mechanism Evaluation in Reference [1] to Farley Units 1

&2 An evaluation of degradation mechanisms that could potentially impact the reliability of the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections was performed for the industry in Reference [1]. Evaluated mechanisms included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG MS and FW nozzles. This observation was acknowledged by the NRC in Section 3.8, page 6, second paragraph of the Reference

[12] Safety Evaluation (SE) for the Vogtle Units 1 & 2 Request for Alternative. Since the materials and operating conditions of the SG feedwater nozzles considered in Reference

[1] are consistent with Farley Units 1 & 2 (per Tables A1 and A2 of Appendix A), the Reference 1 conclusion is also applicable to Farley Units 1 & 2. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in Reference [1].

Applicability of the Stress Analysis in Reference [1] to Farley Units 1 & 2 Finite element analysis (FEA) was performed in Reference [1] to determine the stresses in the SG FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections for representative plants in the industry. The analysis was performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation.

The applicability of the FEA analysis to Farley Units 1 & 2 is shown in Appendix A and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [1] stress analysis are compared to those at Farley Units 1 & 2 in Table 1.

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Table 1 Comparison of Model Geometry in Reference [1] with Farley Units 1 & 2 Modeled in Farley Ri/t Ratio of Component Parameter EPRI Report Units1 and 2 Farley to Model ID (in) 168.88 168.5 NA SG Shell Thickness (in) 3.52 4.21 NA (Ri/t) 24.0 20.0 0.833 ID (in) 16.5 16.51 NA FW Nozzle Thickness (in) 6.0 4.78 NA (Ri/t) 1.375 1.727 1.256 As shown in Table 1, the ratio of the (Ri/t) values at Farley Units 1 & 2 to those of the Reference [1] model is 0.833 for the SG shell and 1.256 for the FW nozzle. As noted by the NRC in Section 3.8.3.1, page 9, third paragraph of the Reference [12] SER for Vogtle, the dominant stress is the pressure stress. Therefore, the ratio of (Ri/t) values determined in Table 1 can be used to scale up the stresses of the Reference [1] report to obtain plant-specific stresses for Farley Units 1 & 2. The Ri/t ratio for the SG shell from Table 1 is applicable to the Item No. C2.21 component since the weld is located in the shell; the Ri/t ratio for the FW nozzle from Table 1 is applicable to Item No. C2.22 since the component is associated with the nozzle.

In selection of the transients in Section 5 of Reference [1] and the subsequent stress analyses in Section 7, test conditions beyond a system leakage test were not considered since pressure tests at Farley Units 1 & 2 are performed at normal operating conditions.

No hydrostatic testing has been performed at Farley Units 1 & 2 since steam generator replacement.

Applicability of the Flaw Tolerance Evaluation in Reference [1] to Farley Units 1 & 2 Flaw tolerance evaluations were performed in Reference [1] consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI) followed by subsequent in-service inspections (ISI), the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year is met. The PFM analysis in Reference [1] was performed using the PRobabilistic OptiMization of InSpEction (PROMISE) Version 1.0 software, developed by Structural Integrity Associates. As part of the NRCs review of SNCs alternative request, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan dated May 14, 2020 (ADAMS Accession No. ML20128J311). The software assumes 100% coverage for the PSI examination.

In Section 8.2.2.2 of Reference [1], a nozzle flaw density of 0.001 flaws per nozzle was assumed for the nozzle inside radius sections. In Section 3.8.5 of the SE for Vogtle [12],

the NRC indicated that a nozzle flaw density of 0.1 flaws per nozzle should have been used. Sensitivity studies performed in Section 8.2.4.3.4 in Reference [1] indicated that by changing the number of flaws in the nozzle inside radius sections from 0.001 to 0.1, the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1x10-6 per year. The NRC determined E-4

Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) that, for the Vogtle alternative, multiplying by a fractional flaw density is a reasonable approach.

Since the sensitivity studies performed in References [1] involve PSI/ISI scenarios that are different from those at Farley Units 1 & 2, supplemental analyses were performed for the plant-specific inspection scenarios as detailed below.

For the Farley Unit 1 replacement SGs, PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacement. The PSI/ISI scenario considered is therefore PSI plus two 10-year ISI examinations to be followed by a 30-year ISI examination (PSI+10+20+50).

For the Farley Unit 2 replacement SGs, PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacement. The inspections in the second 10-year interval following the replacement have not been fully completed since two welds are left to be inspected; therefore, credit was not taken for the second 10-year inspection interval. The PSI/ISI scenario considered is therefore PSI plus one 10-year ISI examination to be followed by a 30-year ISI examination (PSI+10+40).

The PSI/ISI scenario for Farley Unit 2 (PSI+10+40) is used in the evaluation since it is more limiting than that at Farley Unit 1, as it has one less exam following preservice for the inner radius for APR2-3100-IR8R as shown in Appendix B. Evaluations were performed for both the critical nozzle inside radius section and the critical nozzle-to-shell weld location identified in Reference [1].

Item No. C2.22 (nozzle inside radius section) - From Reference [1], the critical location for the inside radius section is FW nozzle Case ID FEW-P1N. An evaluation similar to that shown in Table 8-28 in Reference [1] was performed for this location at Farley assuming a nozzle flaw density of 0.1, a stress multiplier of 1.5 (the same as that used in Reference [1], Table 8-28), a fracture toughness of 200 ksi¥in and a standard deviation 5 ksi¥in, as recommended by the NRC in the Vogtle SE [12]. The fracture toughness of 200 ksi¥in is conservative in application to Farley Units 1 & 2 since the maximum RTNDT values of 10°F for the FW shell and nozzle materials would ensure upper shelf behavior for all transients. It should be noted that the stress multiplier of 1.5 is greater than the plant specific stress multiplier of 1.25 determined in Table 1 for the FW nozzle (the relevant component for C2.22) and is therefore conservative. The evaluation was performed for a period of 80 years from the PSI examination of the replacement SGs.

The results of the evaluation are summarized in Table 2 and show that after 80 years of plant operation from the initial PSI examinations, the probabilities of rupture and leakage remain well below the acceptance criterion of 1.0x10-6.

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Table 2 Plant-Specific Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the Farley Units 1 & 2 Feedwater Nozzle Inside Radius Section (Case ID FEW-P1N from Reference [1])

Probability per Year for Combined Case KIC = 200 ksi¥in.

Time SD = 5 ksi¥in.

(yr) Stress Multiplier = 1.5 Nozzle Flaw Density = 0.1 PSI+10+40 Leak Rupture 10 1.00E-09 1.70E-08 20 5.00E-10 1.30E-08 30 3.33E-10 5.20E-08 40 2.50E-10 2.69E-07 50 2.00E-10 2.18E-07 60 1.67E-10 1.89E-07 70 1.43E-10 1.77E-07 80 1.25E-10 1.76E-07 Item No. C2.21 (nozzle-to-shell weld) - For the feedwater nozzle-to-shell weld, Table 8-16 of Reference [1] indicates that the critical Case ID is FEW-P3A. For the Farley evaluation, a nozzle flaw density of 1 flaw per nozzle was used as accepted by the NRC in the Vogtle SE [12]. A fracture roughness of 200 ksi¥in and standard deviation 5 ksi¥in were also used with a stress multiplier of 1.2. This stress multiplier was chosen such that the probability of rupture or leakage would be as close as possible to the acceptance criteria of 1.0E-06. It should be noted that the stress multiplier of 1.2 is greater than the plant specific stress multiplier of 0.833 determined in Table 1 for the SG shell (the relevant component for C2.21) and therefore conservative. The results of the evaluation are summarized in Table 3 and show that after 80 years of plant operation from the initial PSI examinations, the probabilities of rupture and leakage are below the acceptance criterion of 1.0x10-6.

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Table 3 Plant-Specific Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the Farley Units 1 & 2 Feedwater Nozzle-to-Shell Weld (Case ID FEW-P3A from Reference [1])

Probability per Year for Combined Case KIC = 200 ksi¥in.

Time SD = 5 ksi¥in.

(yr) Stress Multiplier = 1.2 Nozzle Flaw Density = 1 PSI+10+40 Leak Rupture 10 1.00E-08 1.00E-08 20 5.00E-09 5.00E-09 30 3.33E-09 3.33E-08 40 2.50E-09 9.20E-07 50 2.00E-09 7.40E-07 60 1.67E-09 6.32E-07 70 1.43E-09 5.96E-07 80 1.25E-09 8.09E-07 The plant-specific PFM evaluations presented in Tables 2 and 3 for Farley Units 1 & 2 indicate that with conservative inputs of the critical parameters, the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-6. It should be noted that the analyses involve conservative assumptions with regards to the PSI/ISI scenarios. Furthermore, the evaluation [1] was performed for 80 years following the initial PSI examinations, which is longer than the extension being sought by SNC in this Request for Alternative.

The Farley PFM evaluation provided above are within the bounds of the DFM evaluations in Table 8-31 of Reference [1]. This demonstrates that it takes approximately 80 years for a postulated flaw with an initial depth equal to ASME Code,Section XI acceptance standards to grow to a depth where the maximum stress intensity factor (K) exceeds the ASME Code,Section XI allowable fracture toughness.

Inspection History Inspection history for Farley Units 1 & 2 (including examinations performed to date, examination findings, and inspection coverage) is presented in Appendix B. As shown in this appendix, all welds/components have examination coverage greater than 90%

(essentially 100% per IWA-2200 of ASME Code,Section XI). Also, as shown in E-7

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Appendix B, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

The inspection history for these components as obtained from an industry survey is presented in Appendix C. The results of the survey indicate that these components are very flaw tolerant.

Conclusion It is concluded that the SG FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis [1], supplemented by plant-specific evaluations performed as part of this Request for Alternative, demonstrate that using conservative PSI/ISI inspection scenarios, the NRC safety goal of 10-6 failures per reactor year is met with considerable margin for the inspection interval requested in this code alternative. Plant-specific applicability of the technical basis to Farley Units 1 & 2 is demonstrated in Appendix A.

The requested inspection interval provides an acceptable level of quality and safety in lieu of the current ASME Code,Section XI 10-year inspection frequency.

Operating and examination experience demonstrates that these components have performed with very high reliability, mainly due to their robust design. Appendix B shows the ISI examination history of the SG FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections at Farley Units 1 & 2.

No flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations, as shown in Appendix B. In addition, it is important to note all other inspection activities, including the system leakage test (Examination Category C-H) conducted each inservice inspection period (approximately every other refueling outage), will continue to be performed, providing further assurance of safety.

Finally, as discussed in Reference [2], for situations where no active degradation mechanism is present, it was concluded that subsequent ISI examinations do not provide additional value after PSI has been performed and the inspection volumes have been confirmed to be free of defects.

Therefore, SNC requests that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1).

6.0 DURATION OF PROPOSED ALTERNATIVE:

The proposed Alternative is requested for the remainder of the 5th Inservice Inspection through 6th Inspection (ISI) Interval for Farley Units 1 & 2, currently scheduled to end on 11/30/37.

7.0 PRECEDENT

The following previous submittal has been made by SNC to provide relief from the ASME Code,Section XI Examination Category C-B (Item Nos. C2.21 and C2.22) surface and volumetric examinations based on the Reference [1] technical basis report:

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Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) x Letter from C. A. Gayheart (Southern Nuclear) to the U.S. NRC, Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0, dated September 9, 2020, ADAMS Accession No. ML20253A311 [14].

x Letter from Michael T. Markley (USNRC) to Cheryl A. Gayheart (Southern Nuclear), Vogtle Electric Generating Plant, Units 1 & 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code (EPID L-2020-LLR-0109), dated January 11, 2021, ADAMS Accession No. ML20352A155 [12].

In addition, the following is a list of approved actions (including relief requests and topical reports) related to inspections of SG welds and components:

x Letter from J. W. Clifford (NRC) to S. E. Scace (Northeast Nuclear Energy Company), Safety Evaluation of the Relief Request Associated with the First and Second 10-Year Interval of the Inservice Inspection (ISI) Plan, Millstone Nuclear Power Station, Unit 3 (TAC No. MA 5446), dated July 24, 2000, ADAMS Accession No. ML003730922.

x Letter from R. L. Emch (NRC) to J. B. Beasley, Jr. (SNOC), Second 10-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant, Units 1 & 2 (TAC No. MB0603 and MB0604), dated June 20, 2001, ADAMS Accession No. ML011640178.

x Letter from T. H. Boyce (NRC) to C. L. Burton (CP&L), Shearon Harris Nuclear Power Plant Unit 1 - Request for Relief 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, 2R2-011 for the Second Ten-Year Interval Inservice Inspection Program Plan (TAC Nos. ME0609, ME0610, ME0611, ME0612, ME0613, ME0614 and ME0615), dated January 7, 2010, ADAMS Accession No. ML093561419.

x Letter from M, Khanna (NRC) to D. A. Heacock (Dominion Nuclear Connecticut Inc.), Millstone Power Plant Unit No. 2 - Issuance of Relief Requests RR-89-69 Through RR-89-78 Regarding Third 10-Year Interval Inservice Inspection plan (TAC Nos. ME5998 Through ME6006), dated March 12, 2012, ADAMS Accession No. ML120541062.

x Letter from R. J. Pascarelli (NRC) to E. D. Halpin (PG&E), Diablo Canyon Plant, Units 1 & 2 - Relief Request; NDE SG-MS-IR, Main Steam Nozzle Inner Radius Examination Impracticality, Third 10-Year Interval, American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Inservice Inspection Program (CAC Nos. MF6646 and MF6647), dated December 8, 2015, ADAMS Accession No. ML15337A021.

In addition, there are precedents related to similar topical reports that justify relief for Class 1 nozzles:

x Based on studies presented in Reference [3], the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference [4].

x Based on work performed in BWRVIP-108 [5] and BWRVIP-241 [7], the NRC approved the reduction of BWR vessel feedwater nozzle-to-shell weld examinations (Item No. B3.90 for BWRs from 100% to a 25% sample of each nozzle type every 10 years) in References [6] and [8]. The work performed in E-9

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BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 [9], which has been conditionally approved by the NRC in Revision 19 of Regulatory Guide 1.147 [10].

8.0 ACRONYMS

ASME American Society of Mechanical Engineers B&W Babcock and Wilcox BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Program CE Combustion Engineering CFR Code of Federal Regulations DFM Deterministic fracture mechanics EAF Environmentally assisted fatigue EPRI Electric Power Research Institute FAC Flow accelerated corrosion FEA Finite element analysis FW Feedwater ISI Inservice Inspection MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system O.D. Outside diameter PDI Probability of detection PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor SCC Stress corrosion cracking SG Steam Generator SNC Southern Nuclear Company E-10

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9.0 REFERENCES

1. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590.

2. American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR)

Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.

3. B. A. Bishop, C. Boggess, N. Palm, Risk-Informed extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Rev. 3, October 2011.

4. US NRC, Revised Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, Pressurized Water Reactor Owners Group, Project No. 694, July 26, 2011, ADAMS Accession No. ML111600303.

5. BWRVIP-108: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2002. 1003557.

6. US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108), December 19, 2007, ADAMS Accession No. ML073600374.

7. BWRVIP-241: BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005.

8. US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241), April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233.

9. Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR)

Nozzle Inner Radius and Nozzle-to-Shell Welds, ASME Code Section XI, Division 1, Approval Date: February 20, 2004.

10. U. S. NRC Regulatory Guide 1.147, Revision 19, Inservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1, dated October 2019.

11. N. Palm (EPRI), BWR Vessel & Internals Project (BWRVIP) Memo No. 2019-016, White Paper on Suggested Content for PFM Submittals to the NRC, February 27, 2019, ADAMS Accession No. ML19241A545.

12. Letter from Michael T. Markley (USNRC) to Cheryl A. Gayheart (Southern Nuclear),

Vogtle Electric Generating Plant, Units 1 & 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code (EPID L-2020-LLR-0109), dated January 11, 2021, ADAMS Accession No. ML20352A155.

13. Not used.

14. Letter from C. A. Gayheart (Southern Nuclear) to the U.S. NRC, Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0, dated September 9, 2020, ADAMS Accession No. ML20253A311.

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Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

APPENDIX A FARLEY UNITS 1 & 2 APPLICABILITY E-12

Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Section 9, Plant-Specific Applicability of Reference [1] provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Farley Units 1 & 2 is provided in Table A1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report are applicable to Farley Units 1 & 2.

Table A1 Applicability of Reference [1] Representative Analyses to Farley Units 1 & 2 Category Requirement from Reference Applicability to Farley Units 1 & 2

[1]

General The nozzle-to-shell weld shall The Farley Units 1 & 2 FW nozzle Requirements be one of the configurations configurations are shown in Figure shown in Figure 1-1 or Figure A1 and are representative of the 1-2 of Reference [1]. configuration shown in Figure 1-1 of Reference [1].

The materials of the SG shell The Farley Units 1 & 2 FW nozzles and FW nozzles must be low are fabricated of SA-508, Class 3a alloy ferritic steels which material, and the SG shells are conform to the requirements of fabricated from SA-508, Cl. 3 ASME Code,Section XI, material. The RTNDT values are 10°F Appendix G, Paragraph G- for the FW nozzles and 10°F for the 2110. SG shells (so the RTNDT of 60°F used in the EPRI report is bounding).

These materials conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The number of transients The transient cycles in Table 5-5 of shown in Table 5-5 of Reference [1] meet or exceed the 60-Reference [1] are bounding for year projected cycles for Farley Units application over a 60-year 1 & 2 as shown in Table A2 .

operating life.

SG Feedwater The piping attached to the FW The Farley Units 1 & 2 FW piping Nozzle nozzle must be 14-inch to 18- lines are 16-inch NPS.

inch NPS.

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Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Category Requirement from Reference Applicability to Farley Units 1 & 2

[1]

The FW nozzle design must The Farley Units 1 & 2 FW nozzle have an integrally attached configuration has an integrally thermal sleeve attached thermal sleeve.

Auxiliary feedwater nozzles N/A for Farley Units 1 & 2.

connected directly to the SG are not covered in this evaluation.

Figure A1 E-14

Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Table A2 Transient Cycles for Farley Units 1 & 2 in Comparison to the Requirements in Reference

[1]

Transient Cycles From Unit 1 60-Year Unit 2 60-Year Allowable Table 5-5 of Projected Cycles Projected Cycles Cycles From EPRI Report From Table 10 of From Table 11 of Tables 10 and 3002014590 [A1] [A1] 11 of [A1]

[1]

Heatup/Cooldown 300 80/78(5) 63/62(5) 200 Plant Loading(1) 5000 321 282 2000 Plant Unloading(2) 5000 338 486 2000 Loss of Load(3) 360 254 128 760 Loss of Power(4) 60 3 3 40 Notes:

(1) Transient listed as Small Step Load Increase in Tables 10 and 11 of [A1].

(2) Transient listed as Small Step Load Decrease in Tables 10 and 11 of [A1].

(3) Loss of Load transient is a bundled to conservatively envelope a combination of several transients listed in Tables 10 and 11 of [A1]:

x Loss of Load w/o Rx Trip x Loss of RCS Flow 1 Loop x Large Step Load Decrease x Reactor Trip (Cooldown and SI) x Reactor Trip (Cooldown no SI) x Reactor Trip (No Cooldown)

(4) Transient listed as Loss of Offsite Power in Tables 10 and 11 of [A1].

(5) Cycles for Heatup and Cooldown, respectively.

Appendix A References A1. Structural Integrity Associates, Inc. Calculation No, FP-FNP-324, Revision 0, FatiguePro 4 Plant Farley Data Update - through 2021.

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Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

APPENDIX B FARLEY UNITS 1 & 2 INSPECTION HISTORY E-16

Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Date Interval/Period Components ID Exam Coverage Results Item 1R20 3rd/3rd NRI >90%

ALA2-3300-8R No. Spring 2006 C2.21 1R27 4th/3rd NRI >90%

ALA2-3300-8R Fall 2016 2R20 4th/1st NRI >90%

APR2-3100-8R Spring 2010 2R27 5th/1st NRI >90%

APR2-3100-8R Fall 2020 Item 1R20 3rd/3rd NRI >90%

ALA2-3300-IR8R C2.22 Spring 2006 1R25 4th/ 2nd NRI >90%

ALA2-3300-IR8R Fall 2013 2R20 4th/1st NRI >90%

APR2-3100-IR8R Spring 2010 2R25 4th/3rd NRI >90%

APR2-3200-IR8R Fall 2017 E-17

Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

APPENDIX C RESULTS OF INDUSTRY SURVEY E-18

Enclosure to NL-22-0756 Proposed Alternative FNP-ISI-ALT-05-05, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Overall Industry Inspection Summary The results of an industry survey of past inspections of SG MS and FW nozzles are summarized in Section 3 of Reference [1]. Table C1 provides a summary of the combined survey results for Item Nos. C2.22, C2.21, and C2.32(1). The results identify that SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Section examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international BWR and PWR units responded to the survey and provided information representing all PWR plant designs currently in operation in the U.S. This included 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR nuclear steam supply system (NSSS) vendors (i.e., Babcock and Wilcox (B&W), Combustion Engineering (CE), and Westinghouse). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32(1) components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG MS and FW nozzles. Only one PWR examination identified two (2) flaws that exceeded ASME Code Section XI acceptance criteria. The flaws were linear indications of 0.3 and 0.5 in length and were detected in a MS nozzle-to-shell weld using magnetic particle examination techniques. The indications were dispositioned by light grinding.

Table C1 - Summary of Survey Results Number of Number Number of Plant Type Reportable of Units Examinations Indications BWR 27 164 0 PWR 47 563 2 Totals 74 727 2 1 Item No. C2.32 is similar to Item No. C2.21 and was evaluated in the Reference [1] technical basis and included in the industry survey. Farley Units 1 & 2 have not performed any examinations on Item No.

C2.32 components.

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