NL-24-0227, Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)

Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)

ML24185A245
Person / Time
Site:	Farley, Vogtle
Issue date:	07/03/2024
From:	Coleman J Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-24-0227
Download: ML24185A245 (1)

Text

>- Southern Nuclear July 3, 2024 Docket Nos.: 50-348 50-424 50-364 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Regulatory Affairs Joseph M. Farley Nuclear Plant, Units 1 and 2 3535 Colonnade Parkway Birmingham AL 35243 205 992 5000 NL-24-0227 Vogtle Electric Generating Plant, Units 1 and 2 Proposed lnservice Inspection Alternative GEN-ISi-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)

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 GEN-ISi-AL T-2024-03 for Farley Nuclear Plant (FNP) Units 1 and 2 and Vogtle Electric Generating Plant (VEGP) Units 1 and 2. This proposed alternative, described in the Enclosure, would increase the inspection interval of ASME Section XI Table IWB-2500-1 Examination Categories B-B and B-D for item numbers B2.11, B2.12, and B3.110 from every ISi interval to every other interval as described in the Enclosure.

The Enclosure provides the justification for the requested alternative. Attachments 1 and 2 contain FNP Units 1 and 2 and VEGP Units 1 and 2 specific information to support the applicability of the methods in the Enclosure. Attachment 3 provides the results of an industry survey of similar ISi examinations. Attachment 4 provides an evaluation of SNC pressurizer weld inspection coverage as low as 15%.

NRC approval is requested within twelve months of acceptance to support application of the revised schedule to the affected examination.

U.S. Nuclear Regulatory Commission NL-24-0227 Page2 This letter contains no NRC commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.

Respectfully submitted,

~

Jamie M. Coleman Regulatory Affairs Director JMC/dsp/cbg

Enclosure:

Attachments:

Proposed Alternative GEN-ISI-ALT-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

1. Plant-Specific Applicability FNP Units 1 and 2

2. Plant-Specific Applicability VEGP Units 1 and 2

3. Results of Industry Survey

4. Evaluation of 15% Minimum Coverage for SNC Plants Pressurizer Welds cc:

Regional Administrator, Region II NRR Project Manager - Farley, Vogtle 1 & 2 Senior Resident Inspector - Farley, Vogtle 1 & 2 RType: CFA04.054, CVC7000

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Proposed lnservice Inspection Alternative GEN-ISi-AL T-2024-03 for Pressurizer Welds Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class:

Class 1

==

Description:==

Pressurizer vessel head, shell-to-head, and nozzle-to-vessel welds Examination Category: Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 1, Category B-O, full penetration welded nozzles in vessels Item Numbers:

B2.11 - Pressurizer, shell-to-head welds, circumferential B2.12 - Pressurizer, shell-to-head welds, longitudinal B3.110 - Pressurizer, nozzle-to-vessel welds Farley Nuclear Plant Unit 1 (FNP1)

ASME ASME Item No.

Component ID Component Description Category B-B B2.11 ALA 1-21 00-4 PZR Bottom Head to Lower Shell B-B B2.11 ALA 1-2100-7 PZR Upper Shell to Top Head B-B B2.12 ALA 1-2100-1 PZR Lower Shell Long Seam B-B B2.12 ALA 1-2100-3 PZR Upper Shell Long Seam B-D B3.110 ALA 1-2100-9 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-10 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-11 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-12 Spray Nozzle to PZR Top Head B-D B3.110 ALA1-2100-13 Safety Nozzle to PZR Top Head B-D B3.110 ALA1-2100-14 Surge Nozzle to PZR Bottom Head Farley Nuclear Plant Unit 2 (FNP2)

ASME ASME Item No.

Component ID Component Description Cate~orv B-B B2.11 APR 1-2100-4 PZR Bottom Head to Lower Shell B-B B2.11 APR 1-2100-7 PZR Upper Shell to Top Head B-B B2.12 APR 1-2100-1 PZR Lower Shell Long Seam B-B B2.12 APR 1-2100-3 PZR Upper Shell Long Seam B-D B3.110 APR 1-21 00-9 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-10 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-11 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-12 PZR Upper Head to Nozzle Weld El-1

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Farley Nuclear Plant Unit 2 (FNP2)

ASME ASME Item No.

Component ID Component Description Cateqorv B-D B3.110 APR1-2100-13 PZR Upper Head to Nozzle Weld B-D B3.110 APR1-2100-14 PZR Lower Head to Nozzle Weld Vogtle Electric Generating Plant Unit 1 (VEGP1)

ASME ASME Item No.

Component ID Component Description Cateqorv B-B B2.11 11201-V6-002-Upper Head to Upper Shell Weld W01 B-B B2.11 11201-V6-002-Lower Shell to Lower Head wos B-B B2.12 11201-V6-002-Upper Shell Longitudinal Weld W06 B-B B2.12 11201-V6-002-Lower Shell Longitudinal Weld W09 B-D B3.110 11201-V6-002-Upper Head to 6" Relief Nozzle Weld W10 B-D B3.110 11201-V6-002-Upper Head to 6" Safety Nozzle Weld W11 B-D B3.110 11201-V6-002-Upper Head to 6" Safety Nozzle Weld W12 B-D B3.110 11201-V6-002-Upper Head to 6" Safety Nozzle Weld W13 B-D B3.110 11201-V6-002-Upper Head to 4" Safety Nozzle Weld W14 B-D B3.110 11201-V6-002-14" Surge Nozzle Weld to Lower Head W16 Vogtle Electric Generating Plant Unit 2 (VEGP2)

ASME ASME Item No.

Component ID Component Description Category B-B B2.11 11201-V6-002-PZR Bottom Head to Lower Shell W01 B-B B2.11 11201-V6-002-PZR Upper Shell to Top Head wos B-B B2.12 11201-V6-002-PZR Lower Shell Long Seam W06 B-B B2.12 11201-V6-002-PZR Upper Shell Long Seam W09 B-D B3.110 21201-V6-002-Upper Head to 6" Safety Nozzle Weld W10 B-D B3.110 21201-V6-002-Upper Head to 6" Safety Nozzle Weld W11 B-D B3.110 21201-V6-002-Upper Head to 6" Safety Nozzle Weld W12 B-D B3.110 21201-V6-002-Upper Head to 6" Safety Nozzle Weld W13 B-D B3.110 21201-V6-002-Upper Head to 4" Safety Nozzle Weld W14 El-2

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Vogtle Electric Generating Plant Unit 2 (VEGP2)

ASME ASME Item No.

Component ID Component Description Cateqorv B-D B3.110 21201-V6-002-14" Surge Nozzle Weld to Lower Head W16 Note: The applicable portion of the Item No. B2.12 longitudinal seam weld is where it intersects the associated Item No. B2.11 (circumferential) weld.

2.0 APPLICABLE CODE EDITION AND ADDENDA:

FNP1&2 The fifth 10-year inservice inspection (ISi) interval Code of record for FNP1 &2 is the 2007 Edition of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI through the 2008 Addenda, "Rules for lnservice Inspection of Nuclear Power Plant Components." The current fifth 10-year interval start date was December 1, 2017 and will end November 30, 2027.

VEGP1&2 The fourth ISi interval Code of record for VEGP1 &2 is the 2007 Edition of ASME Boiler and Pressure Vessel Code,Section XI, with 2008 Addenda, "Rules for lnservice Inspection of Nuclear Power Plant Components." The current fourth ISi interval start date was May 31, 2017 and will end May 30, 2027.

3.0 APPLICABLE CODE REQUIREMENT:

4.0 ASME Section XI IWB-2500(a), Table IWB-2500-1, examination Categories B-B and B-D require examination of the following Item Nos.:

Item No. B2.11 - Volumetric examination of essentially 100% of both circumferential shell-to-head welds during each inspection interval.

Item No. B2.12 - Volumetric examination of one foot of all longitudinal shell-to-head welds that intersect circumferential welds during the first interval and one foot of one longitudinal shell-to-head weld that intersects a circumferential weld during successive intervals.

Item No. B3.110 - Volumetric examination of essentially 100% of all full penetration nozzle-to-vessel welds during each inspection interval.

REASON FOR REQUEST:

The Electric Power Research Institute (EPRI) performed assessments in Reference [9.1]

of the basis for the ASME Code,Section XI examination requirements specified for the above listed ASME Section XI, Division 1 examination categories for pressurizer welds.

The assessments include a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [9.1]

report concluded that the current ASME Code,Section XI ISi examinations can be El-3

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) deferred for some time with no impact to plant safety. Based on the conclusions of the EPRI report supplemented by plant-specific evaluations contained herein, Southern Nuclear (SNC) is requesting an ISi examination deferral for the subject welds. The Reference [9.1] report was developed consistent with the recommendations provided in EPRl's White Paper on suggested content for PFM submittals [9.12] and NRC Regulatory Guide 1.245 for PFM submittals and associated technical basis [9.13, 9.14].

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

FNP1 For FNP1, SNC is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Item No.

Description Category B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds The pressurizer welds received the required PSI examinations prior to service followed by ISi examinations through the first period of the current fifth inspection interval.

The proposed alternative is to defer the ISi examinations for these Item Nos. for the FNP1 pressurizer from the current ASME Code,Section XI inservice inspection interval requirement to every other interval, for the remainder of the fifth inservice inspection interval and through the sixth inservice inspection interval, which is currently scheduled to end of November 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

FNP2 For FNP2, SNC is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Item No.

Description Category B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds The pressurizer welds received the required PSI examinations prior to service followed by ISi examinations through the first period of the current fifth inspection interval.

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Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

The proposed alternative is to defer the ISi examinations for these Item Nos. for the FNP2 pressurizer from the current ASME Code,Section XI inservice inspection interval requirement to every other interval, for the remainder of the fifth inservice inspection interval and through the sixth inservice inspection interval, which is currently scheduled to end of November 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

VEGP1 For VEGP1, SNC is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Item No.

Description Category B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds The pressurizer welds and components received the required PSI examinations prior to service followed by ISi examinations through the first period of the current fourth inspection interval.

The proposed alternative is to defer the ISi examinations for these Item Nos. for the VEGP1 pressurizer from the current ASME Code,Section XI 10-year requirement to every other interval, for the remainder of the fourth 10-year ISi interval and through the fifth 10-year ISi interval, which is currently scheduled to end on May 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

VEGP2 For VEGP2, SNC is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Item No.

Description Category B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds The pressurizer welds and components received the required PSI examinations prior to service followed by ISi examinations through the first period of the current fourth inspection interval.

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Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

The proposed alternative is to defer the ISi examinations for these Item Nos. for the VEGP2 pressurizer from the current ASME Code,Section XI 10-year requirement to every other interval, for the remainder of the fourth 10-year ISi interval and through the fifth 10-year ISi interval, which is currently scheduled to end on May 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

Technical Basis A summary of the key aspects of the technical basis for this request is summarized below. The applicability of the technical basis to the SNC PWR plants is shown in Attachments 1 and 2.

Applicability of the Degradation Mechanism Evaluation in Reference [9.1] to the SNC PWR Plants An evaluation of degradation mechanisms that could potentially impact the reliability of the pressurizer welds was performed in Reference [9.1 ]. The degradation mechanisms that were evaluated 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 pressurizer welds covered in this request. This observation was acknowledged by the NRC in Section 2, page 3, second paragraph of the Reference [9.16] Safety Evaluation (SE) for Salem Units 1 & 2.

The materials and operating conditions for the plants considered in this Request for Alternative are similar to those in Reference [9.1] and therefore the conclusions of that Report apply to the plants in this Request for Alternative. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in Reference [9.1 ].

As part of the technical basis in Reference [9.1], a comprehensive industry survey involving 7 4 units was conducted to determine the degradation history of these components. The survey reviewed examination results from the start of plant operation.

Most of these plants have operated for over 30 years and in some cases over 40 years.

The results showed that no examinations identified any unknown degradation mechanisms (i.e., mechanisms other than those listed above). Based on this exhaustive industry survey, it is concluded that although the emergence of an unknown degradation mechanism cannot be completely ruled out, the possibility of the occurrence of such an unknown degradation mechanism is highly unlikely.

Applicability of the Stress Analysis in Reference [9.1] to the SNC PWR Plants Finite element analysis (FEA) was performed in Reference [9.1] to determine the stresses in the pressurizer welds covered in this request. The analysis was performed using representative Westinghouse plant 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 the SNC PWR plants is demonstrated in Attachments 1 and 2 and confirms that all plant-specific requirements El-6

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Notes:

are met. In particular, the key geometric parameters used in the Reference [9.1] stress analysis are compared to those of the SNC PWR plants in Tables 1 and 2:

Table 1. Pressurizer Shell Dimensions Shell/Clad Plant Shell ID (in)

Thk (in)

Shell Ri/t EPRI Report (Table 4-4 of 84(1) 3.75 I 0.063(1) 11.2(1)(2) f9.1 l)

FNP1 83.74 3.88 I 0.125 10.8 FNP2 83.74 3.88 I 0.125 10.8 VEGP1 84.31 3.75 I 0.063 11.2 VEGP2 84.31 3.75 I 0.063 11.2

1.

Westinghouse pressurizer dimensions, associated with model for bottom head.

2.

Value from Table 4-4 (based on OD) converted to ID-based value for comparison with the SNC PWR plants.

Table 2. Pressurizer Nozzle Dimensions Surge Nzl Surge Nzl Surge Nzl SRV Nzl ID SRV Nzl SRV Nzl Plant ID (in)

Thk (in)

Ri/t (in)

Thk (in)

Ri/t EPRI Report 12.44(1) 3.27(1) 1 _9(1) 5.625(2) 1.19(2) 2.363(2)

(Table 4-5 of [9.11)

FNP 1 11.5 3.28 1.75 5.33 2.72 0.98 FNP2 11.5 3.28 1.75 5.33 2.72 0.98 VEGP1 12.44 3.28 1.9 5.56 2.72 1.02 VEGP2 12.44 3.28 1.9 5.56 2.72 1.02 Notes:

1.

Westinghouse pressurizer nozzle dimensions, associated with model for bottom head.

2. CE pressurizer nozzle dimensions, associated with model for top head.

As noted by the NRC in Section 5.1, page 7, fourth paragraph of the Salem Safety Evaluation (SE) [9.16], the dominant stress is the pressure stress. Therefore, the variation in the R/t ratio determined in Tables 1 and 2 can be used to scale up the stresses of the Reference [9.1] report to obtain the plant-specific stresses for each unit and component. Since all pressurizer welds at FNP1 &2 and VEGP1 &2 listed in Section 1 are shell welds, Table 1 applies. From this table, the maximum (limiting) inside radius-to-thickness ratio (R/t) of the SNC PWR plants (11.2) for VEGP1/2 is identical to that used in the EPRI report. Therefore, the stresses at FNP1 &2 and VEGP1 &2 are bounded by those considered in the EPRI report. The associated stress ratios for each plant (plant (R/t) ratio divided by that used in the EPRI report) are as follows:

FNP1 &2 pressurizer shell welds stress ratio: (10.8/11.2) = 0.96 VEGP1 &2 pressurizer shell welds stress ratio: (11.2/11.2) = 1.00 In the selection of the transients in Section 5 of Reference [9.1] and the subsequent stress analyses in Section 7, test conditions beyond a system leakage test were not considered since pressure tests at the SNC PWR plants are performed at normal El-7

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) operating conditions. No hydrostatic testing had been performed at the SNC PWR plants since the units went into operation.

Applicability of the Flaw Tolerance Evaluation in Reference [9.1] to the SNC PWR Plants Flaw tolerance evaluations were performed in Reference [9.1] consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. Since the configuration considered in Reference [9.1] is consistent with the Westinghouse pressurizer design, the results of the flaw tolerance evaluation are applicable to the SNC PWR plants. 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 Commission's (NRC's) safety goal of 10-5 failures per year is met.

The PFM analysis in Reference [9.1] was performed using the PRobabilistic OptiMization of lnSpEction (PROMISE) Version 2.0 software, developed by Structural Integrity Associates. As part of the NRC's review of Southern Nuclear's previous alternative request VEGP-ISI-AL T-04-04, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRC's audit plan dated May 14, 2020 (ADAMS Accession No. ML20128J311 ). The PFM analysis in Reference [9.1] was performed using the PROMISE Version 2.0 software which has not been audited by the NRC. The only technical difference between the two versions is that in PROMISE Version 1.0, the user-specified examination coverage is applied to all inspections, whereas in PROMISE Version 2.0, the examination coverage can be specified by the user uniquely for each inspection. In both Versions 1.0 and 2.0, the software assumes 100% coverage for the PSI examination. The NRC staff found the use PROMISE Version 2.0 acceptable in Section 3.1, page 5, fourth paragraph of the Reference [9.16] SE for Salem A comparison of the PSI/ISi scenarios used in the sensitivity studies performed in Reference [9.1] to those at the SNC PWR plants is provided below. The assumption of a 30-year ISi deferral used in the analysis of the below scenarios is conservative compared to the 20-year deferral period being requested for the SNC PWR plants.

FNP1 For the FNP1 pressurizer, PSI examinations have been performed followed by ISi examinations over four 10-year intervals (the unit is currently in its fifth ISi interval). The PSI/ISi scenario considered is therefore (PSl+10+20+30+40+70).

FNP2 For the FNP2 pressurizer, PSI examinations have been performed followed by ISi examinations over four 10-year intervals (the unit is currently in its fifth ISi interval). The PSI/ISi scenario considered is therefore (PSl+10+20+30+40+ 70).

VEGP1 For the VEGP1 pressurizer, PSI examinations have been performed followed by ISi examinations over three 10-year intervals (the unit is currently in its fourth ISi interval).

The PSI/ISi scenario considered is therefore (PSl+10+20+30+60).

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Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

VEGP2 For the VEGP2 pressurizer, PSI examinations have been performed followed by ISi examinations over three 10-year intervals (the unit is currently in its fourth ISi interval).

The PSI/ISi scenario considered is therefore (PSl+10+20+30+60).

Limiting PSI/ISi Scenario The most limiting PSI/ISi scenario for the SNC PWR plants is (PSl+10+20+30+60),

associated with VEGP1 &2, since it involves fewer inspections. This scenario was not specifically considered in the Reference [9.1] PFM evaluations in combination with key variables, as evaluated by the NRC in Section 4.0 (page 6) of the Reference [9.16]

Safety Evaluation. Therefore, a new PFM evaluation was performed for this limiting PSI/ISi scenario using PROMISE Version 2.0, the same version used for the evaluations in the EPRI report [9.1]. The evaluations were performed for the critical Case ID from Reference [9.1] (PRSHC-BW-2C) with a combination of the most dominant parameters (stress and fracture toughness) as identified by the NRC in Section 4.0 (page 6) and Section 10 (page 19) of Reference [9.16]. Since all welds under consideration are shell welds, a flaw density of 1.0 was used in the evaluation. This flaw density value was found acceptable by the NRC in Section 9.6 of Reference [9.16]. A fracture toughness of 200 ksi in with a standard deviation of 5 ksi in was used, as recommended by the NRC in Section 10 (page 19) of Reference [9.16]. A stress multiplier of 2.1 was used in the evaluation. This stress multiplier was conservatively chosen such that probability of rupture or leakage will be close to the acceptance criteria of 1.0x10-5 after 80 years. As discussed above, a bounding stress multiplier of 1.0 can be applied to the FNP1 &2 and VEGP1 &2 pressurizer components and therefore the stress multiplier of 2.1 used in the evaluation is conservative. The results of the evaluation are presented in Table 3.

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Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Table 3. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Weld Flaw Density for 80 Years for the FNP1 &2 and VEGP1 &2 Pressurizer Welds (Case ID PRSHC-BW-2C from Reference [9.1])

Probability per Year for Combined Case Kie = 200 ksi in.,

Time SD = 5 ksiin.

(year)

Stress Multiplier= 2.1 Nozzle Flaw Density = 1 PSl+10+20+30+60 Rupture Leak 10 3.90E-07 l.00E-08 20 3.SSE-07 5.00E-09 30 2.40E-07 3.33E-09 40 l.80E-07 2.S0E-09 50 1.44E-07 2.00E-09 60 l.23E-07 l.67E-09 70 l.06E-07 l.43E-09 80 9.25E-08 l.25E-09 The plant-specific PFM evaluation for the FNP1&2 and VEGP1&2 pressurizer PSI/ISi scenario presented in Table 3 indicates that with conservative inputs of the critical parameters, the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-5 failures per year. The stress multiplier applied in Table 3 is greater than the plant specific stress ratios of 0.96 for FNP1 &2 and 1.0 for VEGP1 &2 determined previously from the geometrical data in Table 1 and therefore the stresses and fracture mechanics evaluations in the Reference [9.1] EPRI report are conservative in application to FNP1 &2 and VEGP1 &2. It should also be noted that the evaluation incorporates conservative assumptions regarding the PSI/ISi scenarios. Furthermore, the evaluation was performed for 80 years, which is longer than the deferral being sought by SNC in this Request for Alternative.

In the PFM evaluations in Reference [9.1], the PVRUF initial flaw size distribution was used. This distribution is applicable to thick vessels and not to relatively thin vessels like pressurizers. This issue was raised by the NRC in RAI No. 4 in Reference [9.17]. In response to this RAI, various initial flaw size distributions were used in a sensitivity study

[9.18] which showed that regardless of which distribution was used, the conclusions of Reference [9.1] remain the same. This was found acceptable by the NRC in Section 9.1, page 15, last paragraph of the SE for Salem [9.16].

An evaluation was performed to show acceptability of the low Kie values at the beginning and ending of the heatup/cooldown transient to address the NRC RAI No. 4 for PSEG in Reference [9.17], using the maximum RT NOT value of 60°F allowed by BTP 5-3 [9.21].

The RT NOT value of 60°F either bounds or is consistent with the values provided in Attachments 1 and 2 for the pressurizer materials at FNP1 &2 and VEGP1 &2, respectively. The evaluation was performed for the most critical Case ID (PRSHC-BW-2C) from Reference [9.1], similar to that performed in Reference [9.18] to respond to RAI

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) 1 in Reference [9.17]. The three flaw sizes evaluated in the PSEG RAI response are also considered. The results are shown Figure 1. As seen in this figure, the calculated applied stress intensity factors are bounded with margin by the corresponding K,c calculated as a function of temperature (based on an RT NDT of 60°F) throughout the transient, for all three flaw depths. The heatup/cooldown transient also includes the leak test; hence, this addresses the adequacy of the temperature of the leak test to ensure that the applied stress intensity factor is below the fracture toughness during the leak test.

350 O \\C1AOJ9d:1'\\ZI00119-R!l!Support\\Salem-4031KT\\P2C-RTNOT60-ca20_ptt ]QQ 300 I Temoer.1Lure I 600 250 500 400 300 100 200 100 o.__._ _

_,_____..__,_____,o 0

10000 20000 30000 Time(sec) 40000 50000 G:'

~

8.

E

~

Figure 1. Applied K vs. Fracture Toughness as a Function of Temperature for Case ID PRSHC-BW-2C (RT NDT = 60°F)

The confirmatory DFM evaluation in Table 8-4 of Reference [9.1] provides verification of the above PFM results for the SNC PWR plants by demonstrating 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 As described in Section 8.3.4.1 of Reference [9.1], preservice examination (PSI) refers to the superset of the examinations required by ASME Code, Section Ill during fabrication and required by ASME Code,Section XI prior to service. The Section Ill fabrication examinations required for these components were robust, and any Section XI preservice examinations further contributed to thorough initial examinations.

FNP1 Inspection history for FNP1 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 1. As shown in the attachment, the minimum examination coverage of all welds/components was 75%. Acceptability of inspection coverage as low as 15% for the SNC pressurizer welds is demonstrated in Attachment 4. As shown in Attachment 1, no flaws that El-11

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

FNP2 Inspection history for FNP2 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 1. As shown in the attachment, the minimum examination coverage of all welds/components was 45%. Acceptability of inspection coverage as low as 15% for the SNC pressurizer welds is demonstrated in Attachment 4. As shown in Attachment 1, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

VEGP1 Inspection history for VEGP1 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 2. As shown in the attachment, all welds have coverage greater than 44% with the exception of one weld with minimum examination coverage of 15%. Acceptability of inspection coverage as low as 15% for the SNC pressurizer welds is demonstrated in Attachment

4. As shown in Attachment 2, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

VEGP2 Inspection history for VEGP2 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 2. As shown in the attachment, all welds have coverage greater than 49% with the exception of two welds with a minimum examination coverage of 15%. Acceptability of inspection coverage as low as 15% for the SNC pressurizer welds is demonstrated in Attachment

4. As shown in Attachment 2, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

Industry Survey The inspection history for these components as obtained from an industry survey is presented in Attachment 3. The results of the survey indicate that these components are very flaw tolerant.

Performance Monitoring Examinations at other inspection intervals being requested by SNC for FNP Units 1 and 2 and VEGP Units 1 and 2 is consistent with the previous SNC request for FNP Units 1 and 2 SG feedwater nozzle components [9.19] which was found acceptable by the NRC

[9.20] to provide adequate performance monitoring.

Future Examinations Future interval examinations will be performed in accordance with the ASME inspection and evaluation requirements at the time of inspection. Scope expansion will be performed in accordance with the ASME Section XI code of record. Current rulemaking El-12

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) is in progress to adopt ASME Code Case N-921. Following the current inspection interval, both Farley and Vogtle will update to at least the 2019 version of Section XI as this is the latest version incorporated by reference into 10 CFR 50.55a. Code Case N-921 allows for inspection intervals to be 12 years instead of the traditional 10 years. The below tables provide the schedule of future examinations without and with the endorsement of ASME Code Case N-921. For scheduling purposes, SNC will utilize ASME Code Case N-921 for future examinations if incorporated into the Regulatory Guide 1.147 and incorporated by reference into 10 CFR 50.55a. If ASME Code Case N-921 is not incorporated into the Regulatory Guide 1.147 and incorporated by reference into 10 CFR 50.55a, SNC will utilize the examination window without Code Case N-921.

Previous and Next Scheduled Examination for Each Component FNP1 &2 Item Date Interval Components Next Next Next No.

/ Period ID Exam-Examination Examination ination Window Window Interval without with Code I

Code Case Case N-921 Period N-921 B2.11 1R27 4th/3rd 6th/3rd 12/1/2034 -

12/1/2035 -

10/6/2016 ALA1-2100-4 11/30/2037 11/30/2039 1R29 5th/1st 7th/1st 12/1/2037 -

12/1/2039 -

10/7/2019 ALA1-2100-7 11/30/2040 11/30/2043 2R23 4th/2nd 6th/2nd 12/1/2030 -

12/1/2031 -

10/23/2014 APR 1-2100-4 11/30/2034 11/30/2035 2R26 5th/1st 7th/1st 12/1/2037 -

12/1/3039 -

4/16/2019 APR 1-2100-7 11/30/2040 11/30/2043 B2.12 1R27 4th/3rd ALA1-2100-1 6th/3rd 12/1/2034 -

12/1/2035 -

10/6/2016 11/30/2037 11/30/2039 1R29 5th/1st ALA 1-2100-3 7th/1 St 12/1/2037 -

12/1/2039 -

10/10/2019 11/30/2040 11/30/2043 2R25 4th/3rd 6th/3rd 12/1/2034 -

12/1/2039 -

10/23/2017 APR 1-2100-1 11/30/2037 11/30/2043 2R26 5th/1 St 7th/1 St 12/1/2040 -

12/1/2043 -

4/18/2023 APR 1-2100-3 11/30/2044 11/30/2047 B3.110 1R27 4th/3rd ALA 1-2100-10 6th/3rd 12/1/2034 -

12/1/2035 -

10/8/2016 El-13

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) 1R24 4th/2nd ALA1-2100-11 4/6/2012 1R24 4th/2nd ALA1-2100-12 4/6/2012 1R24 4th/2nd ALA1-2100-13 4/2/2012 1R30 5th/1st ALA1-2100-14 4/1/2021 1R28 5th/1 St ALA 1-2100-9 3/18/2018 2R29 5th/2nd 10/10/2023 APR 1-2100-9 2R29 5th/2nd 10/10/2023 APR 1-2100-10 2R24 4th/3rd 4/20/2016 APR1-2100-11 2R24 4th3rd 4/20/2016 APR1-2100-12 2R26 5th/1st 4/12/2019 APR1-2100-13 2R26 5th/1 St 4/12/2019 APR1-2100-14 El-14 11/30/2037 11/30/2039 6th/3rd 12/1/2030 -

12/1/2031 -

11/30/2034 11/30/2035 6th/2nd 12/1/2030 -

12/1/2031 -

11/30/2034 11/30/2035 5th/2nd 12/1/2030 -

12/1/2031 -

11/30/2034 11/30/2035 7th/1 St 12/ 1/2037 -

12/ 1/3039 -

11/30/2040 11/30/2043 7th/1st 12/ 1/2037 -

12/ 1/3039 -

11/30/2040 11/30/2043 7th/2nd 12/ 1/2040 -

12/ 1/2043 -

11/30/2044 11/30/2047 7th/2nd 12/ 1/2040 -

12/ 1/2043 -

11/30/2044 11/30/2047 6th/3rd 12/1/2034 -

12/1/2035 -

11/30/2037 11/30/2039 6th3rd 12/1/2034 -

12/1/2035 -

11/30/2037 11/30/2039 7th/1st 12/1/2037 -

12/1/2039 -

11/30/2040 11/30/2043 7th/1 St 12/1/2037 -

12/1/2039 -

11/30/2040 11/30/2043

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Next Scheduled Examination for Each Component VEGP1 &2 Item Date Interval Components Next Next No.

/ Period ID Exam-Examination ination Window Interval without I

Code Case Period N-921 B2.11 1 R21 4th/1st 11201-V6-6th/1st 5/31/2037 -

9/25/2018 002-W01 5/30/2040 1R24 4th/2nd 11201-V6-6th/2nd 5/31/2040 -

3/9/2023 002-W0S 5/30/2043 2R22 4th/2nd 21201-V6-5th/2nd 5/31/2040 -

3/11/2022 002-W01 5/30/2043 2R20 4th/1 St 21201-V6-6th/1 St 5/31/2037 -

3/17/2019 002-W0S 3/30/2040 B2.12 1 R21 4th/1 St 11201-V6-6th/1 St 5/31/2037 -

9/25/2018 002-W06 5/30/2040 1 R19 3rd/3rd 11201-V6-5th/3rd 5/31/2034 -

9/23/2015 002-W09 5/30/2037 2R20 4th/1 St 21201-V6-6th/1 St 5/31/2037 -

3/13/2019 002-W06 5/30/2040 2R18 3rd/3rd 21201-V6-5th/3rd 5/31/2034 -

3/10/2016 002-W09 5/30/2037 B3.110 1 R21 4th/1 St 11201-V6-5th/1st 5/31/2037 -

9/20/2018 002-W10 5/30/2040 1 R21 4th/1 st 11201-V6-6th/1 St 5/31/2037 -

9/20/2018 002-W11 5/30/2040 1R23 4th/2nd 11201-V6-6th/2nd 5/31/2040 -

9/21/2021 002-W12 5/30/2043 1R23 4th/2nd 11201-V6-5th/2nd 5/31/2040 -

9/17/2021 002-W13 5/30/2043 1 R19 3rd/3rd 11201-V6-5th/3rd 5/31/2034 -

9/24/2015 002-W14 5/30/2037 El-15 Next Examination Window with Code Case N-921 5/31/2039-5/30/2043 5/31/2043-5/30/2047 5/31/2043-5/30/2047 5/31/2039-5/30/2043 5/31/2039-5/30/2043 5/31/2035-5/30/2039 5/31/2039-5/30/2043 5/31/2035-5/30/2039 5/31/2039-5/30/2043 5/31/2039-5/30/2043 5/31/2043-5/30/2047 5/31/2043-5/30/2047 5/31/2035-5/30/2039

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) 1 R19 3rd/3rd 11201-V5-9/30/2015 002-W15 2R20 4th/1 St 21201-V5-3/12/2019 002-W10 2R20 4th/1st 21201-V5-3/12/2019 002-W11 2R22 4th/2nd 21201-V5-3/14/2022 002-W12 2R22 4th/2nd 21201-V5-2/10/2022 002-W13 2R15 3rd/2nd 21201-V5-9/22/2011 002-W14 2R14 3rd/2nd 21201-V5-3/13/2010 002-W15 Conclusion 5th/3rd 5/31/2034 -

5/31/2035-5/30/2037 5/30/2039 5th/1 St 5/31/2037 -

5/31/2039-5/30/2040 5/30/2043 5th/1st 5/31/2037 -

5/31/2039-5/30/2040 5/30/2043 5th/2nd 5/31/2040 -

5/31/2043-5/30/2043 5/30/2047 5th/2nd 5/31/2040 -

5/31/2043-5/30/2043 5/30/2047 5th/2nd 5/31/2030 -

5/31/2031 -

5/30/2034 5/30/2035 5th/2nd 5/31/2030 -

5/31/2031 -

5/30/2034 5/30/2035 It is concluded that the pressurizer pressure-retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis report [9.1], supplemented by plant-specific evaluations performed as part of this Request for Alternative, demonstrate that using conservative PSI/ISi inspection scenarios for all plants, the NRC safety goal of 10-5 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to the SNC PWR plants is demonstrated in Attachments 1 and 2. The requested ISi deferrals provide 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. Attachments 1 and 2 show the examination history for the pressurizer welds examined in the two most recent 10-year inspection intervals.

In addition to the required PSI examinations for these pressurizer welds, the SNC PWR plants have performed multiple ISi examinations through the current 10-year inspection interval at each plant.

No flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations, as shown in Attachments 1 and 2.

Finally, as discussed in Reference [9.3], 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.

El-16

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

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

6.0 DURATION OF PROPOSED ALTERNATIVE:

FNP1 The proposed alternative is to defer the ISi examinations for these Item Nos. for the FNP1 pressurizer from the current ASME Code,Section XI current ISi interval requirement to every other interval, for the remainder of the fifth inspection interval and through the sixth inspection interval, which is currently scheduled to end on November 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

The proposed alternative is to defer the ISi examinations for these Item Nos. for the FNP2 pressurizer from the current ASME Code,Section XI current ISi interval requirement to every other interval, 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 November 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

VEGP1 The proposed alternative is to defer the ISi examinations for these Item Nos. for the VEGP1 pressurizer from the current ASME Code,Section XI current ISi interval requirement to every other interval, for the remainder of the fourth 10-year ISi interval and through the fifth 10-year ISi interval, which is currently scheduled to end on May 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

VEGP2 The proposed alternative is to defer the ISi examinations for these Item Nos. for the VEGP2 pressurizer from the current ASME Code,Section XI current ISi interval requirement to every other interval, for the remainder of the fourth 10-year ISi interval and through the fifth 10-year ISi interval, which is currently scheduled to end on May 30, 2037. All exams will be scheduled to occur in the same period as the last examination, but with a two-interval inspection periodicity.

7.0 PRECEDENTS

The following previous submittal has been made by PSEG Nuclear to provide relief from the ASME Code,Section XI, Examination Category B-B (Item Nos. B2.11 and B2.12) and Category B-D (Item No. B3.110) surface and volumetric examinations based on the Reference 9.1 technical basis report:

Letter from Paul R. Duke, Jr. (PSEG Nuclear) to USNRC, "Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number El-17

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

B2.11 and B2.12," dated August 5, 2020, ADAMS Accession No. ML20218A587

[9.15].

The USN RC issued a safety evaluation of the PSEG Nuclear request for alternative on April 12, 2021.

Letter from James G. Danna (USNRC) to Eric. Carr (PSEG Nuclear), "Salem Generating Station Unit Nos. 1 and 2 - Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200 (EPID L-2020-LLR-0103)," dated April 12, 2021, ADAMS Accession No. ML20218A587 [9.16].

The following is a list of other Relief Requests and other precedents related to inspections of pressurizer welds and components:

Letter from M. G. Kowal (NRC) to M.A. Balduzi (Entergy Nuclear Operations, Inc.), "Indian Point Nuclear Generating Unit No. 2 - Relief Request No. RR-01 (TAC No. MD4695)," dated September 5, 2007, ADAMS Accession No. ML072130487.

Letter from T. L. Tate (NRC) to Vice President, Operations (Entergy Nuclear Operations, Inc.), "Indian Point Nuclear Generating Unit No. 2 - Safety Evaluation for Relief Request No. IP2-ISI-RR-01 (CAC No. MF082)," dated September 14, 2016, ADAMS Accession No. ML16179A178.

Letter from H. K. Chernoff (NRC) to D. A. Heacock (Dominion Nuclear Connecticut, Inc.), "Millstone Power Station Unit No. 3 - Issuance of Relief Request IR-2-51 through IR-2-60 Regarding Second 10-Year Interval lnservice Inspection Program Plan (TAC Nos. ME3809 through ME3818)," dated April 26, 2011, ADAMS Accession No. ML110691154.

Letter from R. L. Emch (NRC) to J. B. Beasley Jr. (Southern Nuclear Operating Company, Inc.), "Second Ten-Year Interval lnservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant Units 1 and 2 (TAC No. MB0603 and MB0604)," dated June 20, 2001, ADAMS Accession No. ML011640178.

Letter from N. Di Francesco (NRC) to M. J. Pacilio (Exelon Nuclear), Braidwood Station Units 1 and 2 - Relief from Requirements of the ASME Code for the Third 10-Year Interval of lnservice Inspection (TAC Nos. ME9748 and ME9749)," dated January 30, 2013, ADAMS Accession No. ML13016A515.

Letter from E. C. Marinos (NRC) to D. Jamil (Duke Power Company LLC)),

"Catawba Nuclear Station, Unit 1 - Request for Relief 05-CN-004, Limited Weld Examinations During End-of-Cycle 15 Refueling Outage (TAC Nos. MC8337, MC9171, MC9172, MC9173, MC917 4, MC9175, MC9176, MC9177, MC9178, and MC9179)," dated September 25, 2006, ADAMS Accession No. ML062390020.

Letter from J. Boska (NRC) to K. Henderson (Duke Energy Carolinas, LLC)),

"Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 11-CN-001 for the Third 10-Year lnservice Inspection Interval (TAC Nos. ME7277, ME7278, ME7279, ME7280, ME7281, ME7282, AND ME7283)," dated August 20, 2012, ADAMS Accession No. ML12228A723.

Letter from R. J. Pascarelli (NRC) to K. Henderson (Duke Energy Carolinas, LLC)), "Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 14-CN-001, American Society of Mechanical Engineers (ASME)Section XI El-18

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Volumetric Examination Requirements (TAC Nos. MF3527 AND MF3528)," dated October 30, 2014, ADAMS Accession No. ML14295A532.

Letter from R. T. Repko (Duke Energy Carolinas, LLC) to NRC, "Duke Energy Carolinas, LLC (Duke Energy), McGuire Nuclear Station Units 1 and 2, Docket Nos. 50-369 and 50-370, Relief Request Serial# 11-MN-001, Limited Weld Examinations for Refueling Outage 1 EOC20 and 2EOC19," dated September 21, 2011, ADAMS Accession No. ML11279A035.

Letter from J. A. Price (Dominion Nuclear Connecticut, Inc.) to NRC, "Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3, ASME Section XI lnservice Inspection Program, Relief Requests for Limited Coverage Examinations Performed in the Second 10-Year Inspection Interval," dated April 19, 2010, ADAMS Accession No. ML101130187.

Letter from D. H. Corlett (Progress Energy) to NRC, "Shearon Harris Nuclear Power Plant, Unit No. 1, Docket No. 50-400/License No. NPF-63, Second Ten Year Interval lnservice Inspection Program - Final Documentation Including Requests for Relief in Accordance with 10 CFR 50.55a," dated February 5, 2009, ADAMS Accession No. ML090540055.

Letter from D. H. Corlett (Progress Energy) to NRC, "Shearon Harris Nuclear Power Plant, Unit No. 1, Docket No. 50-400/Renewed License No. NPF-63, Response to Request for Additional information Regarding Relief Requests 2R1-018, 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, AND 2R2-011 for the Second Ten Year Interval Inspection Program (TAC Nos. ME0608, ME0609, ME0610, ME0166, ME0612, ME0613, ME0614, AND ME0615)," dated September 24, 2009, ADAMS Accession No. ML092740063.

In addition, other studies have been performed by the industry to extend the inspection interval for various components and have been accepted by the NRC.

Based on studies presented in Reference [9.4], the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference [9.5].

Based on work performed in BWRVIP-108 [9.6] and BWRVIP-241 [9.8], the NRC approved the reduction of BWR vessel 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 [9.7] and [9.9]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 [9.10],

which has been conditionally approved by the NRC in Revision 18 of Regulatory Guide 1.147 [9.11].

8.0 ACRONYMS

ASME B&W BWR BWRVIP CE CFR DFM EAF EPRI FAC American Society of Mechanical Engineers Babcock and Wilcox Boiling Water Reactor Boiling Water Reactor Vessel and Internals Program Combustion Engineering Code of Federal Regulations Deterministic fracture mechanics Environmentally assisted fatigue Electric Power Research Institute Flow accelerated corrosion El-19

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

FEA FW ISi MIC MS NPS NRC NSSS 0.0.

POI PFM PSI PWR PZR sec WEC Finite element analysis Feedwater lnservice Inspection Microbiologically influenced corrosion Main Steam Nominal pipe size Nuclear Regulatory Commission Nuclear steam supply system Outside diameter Probability of detection Probabilistic fracture mechanics Preservice inspection Pressurized Water Reactor Pressurizer Stress corrosion cracking Westinghouse Electric Company

9.0 REFERENCES

9.1 Technical Bases for Inspection Requirements for PWR Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, EPRI, Palo Alto, CA: 2019. 3002015905.

9.2 Not used.

9.3 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.

9.4 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.

9.5 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.

9.6 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.

9.7 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.

9.8 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.

9.9 US NRC, Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling El-20

Enclosure to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241 ),"

April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233.

9.10 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.

9.11 U. S. NRC Regulatory Guide 1.147, Revision 18, "lnservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1," dated March 2017.

9.12 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.

9.13 USNRC Regulatory Guide 1.245, Revision 0, "Preparing Probabilistic Fracture Mechanics Submittals," January 2022.

9.14 USN RC Report NUREG/CR-7278, "Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications," January 2022.

9.15 Letter from Paul R. Duke, Jr. (PSEG Nuclear) to USNRC, "Proposed Alternative for Examination of ASME Section XI, Examination Category B-8, Item Number 82.11 and 82.12," dated August 5, 2020, ADAMS Accession No. ML20218A587.

9.16 Letter from James G. Danna (USNRC) to Eric Carr (PSEG Nuclear), "Salem Generating Station Unit Nos. 1 and 2 -Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200 (EPID L-2020-LLR-0103)," dated April 12, 2021, ADAMS Accession No. ML20218A587.

9.17 Letter from James Kim (USN RC) to Paul R. Duke, Jr. (PSEG Nuclear), "Requests for Additional Information Regarding Salem Generating Station Units Nos. 1 and 2 Regarding Alternative for Examination of ASME Section XI, Category B-8, Item Number 82.11 and 82.12, EPID L-2020-LRR-0103," dated February 11, 2021, ADAMS Accession No. ML21043A144.

9.18 Letter from Paul R. Duke, Jr. (PSEG Nuclear) to USNRC, "Response to Request for Additional Information for Proposed Alternative for Examination of ASME Section XI, Examination Category B-8, Item Number 82.11 and 82.12," dated April 12, 2021, ADAMS Accession No. ML21102A024.

9.19 Letter NL-22-0756 from C. H. Gayheart (Southern Nuclear) to US. NRC. "Joseph M. Farley Nuclear Plant - Units 1 and 2 Proposed lnservice Inspection Alternative FNP-ISI-AL T-05-05, Version 1.0," dated September 30, 2022, ADAMS Accession No. ML22273A159.

9.20 Letter form J. M. Heisserer (U S. NRC) to J. M. Coleman (Southern Nuclear Operating Co. Inc) "Joseph M. Farley Nuclear Plant, Units 1 And 2 - Proposed lnservice Inspection Alternative FNP-ISI-AL T-05-05, Version 1.0, to the Requirements of the ASME Code (EPID L-2022-LLR-0068)," dated August 30, 2023, ADAMS Accession No. ML23164A120.

9.21 NUREG-0800 - Chapter 5, Branch Technical Position (BTP) 5-3, Revision 2, Fracture Toughness Requirements.

El-21

ATTACHMENT 1 PLANT-SPECIFIC APPLICABILITY FNP UNITS 1 AND 2 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Section 9 of Reference [1-1] 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 FNP1 &2 is provided in Table 1-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report are applicable to FNP1 &2.

Table 1-1 Applicability of Reference [1-1] Representative Analyses to FNP1&2 Pressurizer Surge Nozzle and Bottom Head Welds (Item Nos. B2.11 1 B2.12 and B3.110)

Category Requirement from Reference [1-1]

Applicability to FNP1 &2 General The plant-specific pressurizer general transients As shown in Table 1-3, the FNP1&2 Requirements and cycles must be bounded by those shown in general transients are bounded by the Table 5-6 for a 60-year operating life. It should transients listed in Table 5-6 of be noted that the number of cycles were Reference (1-1].

extrapolated to 80 years in the evaluations.

The materials of the pressurizer shell and surge The FN P1 &2 pressurizer shell is nozzle must be low alloy ferritic steels which fabricated from SA-533 Gr A Cl 2 conform to the requirements of ASME Code, material and the surge nozzle isSection XI, Appendix G, Paragraph G-2110.

fabricated from SA-508 Cl 2 material.

The associated RTNoT values are 10°F or less, which is bounded by those used in Reference (1-1].

Both materials are low alloy ferritic steels that conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific pressurizer surge nozzle and The FN P1 &2 weld configurations are Requirements bottom head weld configurations must conform shown in Figures 1-1 and 1-2 and to those shown in Figure 1-1 (Item No. B2.11 ),

show conformance with the Figures Figure 1-2 (Item No. B2.12) and Figures 1-4, 1-shown in Reference (1-1 ].

5 or 1-6 (Item No. B3.110) of Reference (1-1 ].

The plant-specific dimensions of the pressurizer As shown in Table 1-2, the FNP1&2 shell and the surge nozzle must be within the pressurizer shell and surge nozzle range of values listed in Table 9-1 of Reference dimensions are within the range of (1-1 ].

values listed in Table 9-1 of Reference (1-1 l.

Al-1 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Category Requirement from Reference [1-1]

The plant-specific lnsurge/Outsurge transient definitions (temperature difference between the pressurizer shell and the pressurizer surge nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference (1-1].

Applicability to FNP1 &2 As shown in Table 1-4, the FNP1&2 lnsurge/Outsurge transients are bounded by the transients listed in Table 5-10 of Reference (1-1].

Pressurizer Top Head Welds

{Item Nos. B2.11, B2.12 and B3.110)

Category Requirement from Reference [1-1]

Applicability to FNP1 &2 General The plant-specific pressurizer general transients As shown in Table 1-3, the FNP1&2 Requirements and cycles must be bounded by those shown in general transients are bounded by the Table 5-6 of Reference (1-1] for a 60-year transients listed in Table 5-6 of operating life. It should be noted that the Reference (1-1].

number of cycles were extrapolated to 80 years in the evaluations.

The materials of the pressurizer shell and top The FN P1 &2 pressurizer shell is head nozzles must be low alloy ferritic steels fabricated from SA-533 Gr A Cl 2 which conform to the requirements of ASME material and the top head nozzles are Code,Section XI, Appendix G, Paragraph G-fabricated from SA-508 Cl 2/2a 2110.

material. The associated RT NOT values are 10°F, except for the one safety nozzle fabricated of SA-508 Cl 2a material, which has an RTNoT value of 60°F. These values are bounded by those used in Reference (1-1].

All these materials are low alloy ferritic steels that conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific pressurizer top head nozzle The FN P1 &2 top head nozzle weld Requirements weld configurations must conform to those configurations are shown in Figures 1-shown in Figure 1-1 (Item No. B2.11), Figure 1-3 and 1-4 and conform to the Figures 2 (Item No. B2.12) and Figures 1-4, 1-5 or 1-6 shown in Reference (1-1 ].

(Item No. B3.110) of Reference (1-1 ].

Al-2 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Category Requirement from Reference [1-1]

Applicability to FNP1 &2 The plant-specific dimensions of the pressurizer As shown in Table 1-2, the FNP1&2 shell and the top head nozzles must be within pressurizer shell and top head nozzle the range of values listed in Table 9-1 of dimensions are within the range of Reference [1-1].

values listed in Table 9-1 of Reference

[1-1 l.

Table 1-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with FNP1&2 Component Geometric Parameter For a Westinghouse Plant FNP1&2 Dimensions Pressurizer Inside Diameter (in)

Must be between 80 and 88 84" Shell NPS of piping or component (e.g.,

Surge Nozzle reducer) attached to nozzle safe-end Must be between 12 and 18 14" (in) (1)

Safety/Relief NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end Must be between 4 and 8 6"

Nozzle (in) (1)

NPS of piping or component (e.g.,

Spray Nozzle reducer) attached to nozzle safe-end Must be between 4 and 6 4"

(in) (1)

Note:

(1) Depending on the plant-specific configuration, the NPS of the piping (or component) attached directly to the nozzle (i.e., no safe-end) or to a safe-end extension may be used.

Al-3 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Table 1-3 Comparison of FNP1&2 General Transients to Requirements in Reference [1-1]

Number of Cycles for 60 FNP1 FNP2 Transient Years from 60-Year 60-Year Table 5-6 of Projection Projection Reference [1-1]

Heatup /

300 80 I 79 64 I 63 Cooldown Loss of Load (Large Step Load Decrease, Loss of 360 254 125 Power, Loss of Flow, Reactor Trip)

Table 1-4 Comparison of FNP1&2 lnsurge/Outsurge Transients to Requirements in Reference [1-1]

60-Year No. of Cycles From Table 5-10 of FNP1 FNP2

.LH (DF)(1)

Reference [1-1] (For 60-Year 60-Year Westinghouse and CE Projection(2l Projection(2l Plants) 330 600 0

0 320 3,000 101 104 103 1,500 146 276 Notes:

(1) b.T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

(2) Table 15 of Reference [1-2].

Al-4 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Table 1-5 FNP1 Inspection History Item Component Exam Interval/Period/

Exam No.

ID Date Outage Results B2.11 ALA 1-21 00-4 10/6/2016 4/3/1 R27 A

B2.11 ALA 1-21 00-4 4/13/2006 3/3/1 R20 A

B2.11 ALA1-2100-7 10/7/2019 5/1/1 R29 A

B2.11 ALA1-2100-7 10/21/2010 4/1/1 R23 A

B2.12 ALA1-2100-1 10/6/2016 4/3/1 R27 A

B2.12 ALA1-2100-1 4/13/2006 3/3/1 R20 A

B2.12 ALA1-2100-3 10/7/2019 5/1/1 R29 A

B2.12 ALA1-2100-3 10/21/2010 4/1/1 R23 A

B3.110 ALA1-2100-9 4/11/2018 5/1/1 R28 A

B3.110 ALA1-2100-9 4/13/2009 4/1/1 R22 A

B3.110 ALA 1-2100-10/8/2016 4/3/1 R27 A

10 B3.110 ALA 1-2100-10/30/2007 3/3/1 R21 A

10 B3.110 ALA 1-2100-4/6/2012 4/2/1 R24 A

11 B3.110 ALA 1-2100-10/16/2001 3/2/1R17 A

11 B3.110 ALA 1-2100-4/6/2012 4/2/1 R24 A

12 B3.110 ALA 1-2100-10/17/2001 3/2/1R17 A

12 B3.110 ALA 1-2100-4/6/2012 4/2/1 R24 A

13 B3.110 ALA 1-2100-10/17/2001 3/2/1R17 A

13 B3.110 ALA 1-2100-4/2/2021 5/1/1 R30 A

14 Al-5 Coverage Relief Request

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A 78.6%

Yes 75%

Yes 78.6%

Yes 75%

Yes 78.6 Yes 75%

Yes 75.2%

Yes to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Item Component Exam Interval/Period/

No.

ID Date Outage B3.110 ALA 1-2100-4/10/2012 4/2/1 R24 14 A = Acceptable Table 1-6 Exam Results A

FNP2 Inspection History Item Component Exam Date Interval/Period/

Exam No.

ID Outage Results B2.11 APR1-2100-4 10/20/2014 4/2/2R23 A

B2.11 APR1-2100-4 10/24/2005 3/2/2R17 A

B2.11 APR1-2100-7 4/16/2019 5/1/2R26 A

B2.11 APR1-2100-7 10/26/2008 4/1/2R19 A

B2.12 APR1-2100-1 10/21/2017 4/3/2R25 A

B2.12 APR1-2100-1 3/19/2004 3/1/2R16 A

B2.12 APR1-2100-3 4/15/2019 5/1/2R26 A

B2.12 APR1-2100-3 10/26/2008 4/1/2R19 A

B3.110 APR1-2100-9 10/20/2014 4/2/2R23 A

B3.110 APR1-2100-9 9/24/2002 3/2/2R15 A

B3.110 APR1-2100-10/20/2014 4/2/2R23 A

10 B3.110 APR1-2100-9/24/2002 3/2/2R15 A

10 B3.110 APR1-2100-4/14/2016 4/3/2R24 A

11 B3.110 APR1-2100-10/19/2005 3/3/2R17 A

11 B3.110 APR1-2100-4/14/2016 4/3/2R24 A

12 B3.110 APR1-2100-10/19/2005 3/3/2R17 A

12 Al-6 Coverage Relief Request 75.2%

Yes Coverage Relief Request

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A

>90%

N/A 100%

N/A

>90%

N/A

>90%

N/A 61.1%

Yes 75%

Yes 61.1%

Yes 75%

Yes 50%

Yes 50%

Yes 50%

Yes 50%

Yes to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Item Component Exam Date No.

ID B3.110 APR1-2100-4/12/2019 13 B3.110 APR1-2100-10/26/2008 13 B3.110 APR1-2100-4/12/2019 14 B3.110 APR1-2100-11/4/2008 14 A= Acceptable Interval/Period/

Exam Coverage Relief Outage Results Request 5/1/2R26 A

65%

Yes 4/1/2R19 A

75%

Yes 5/1/2R26 A

45%

Yes 4/1/2R19 A

60%

Yes Al-7 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Figure 1-1. FNP1 &2 Pressurizer Vessel

1. 3 f

01B13K001 91,50" 0,D, 288" CIRC.

-.,.. t"" -

L_J 32,00" 87.00".D.

_l 273" CIRC, Al-8

1. 7

1. 6

1. 2

1. 5 s

1. 8 SUPPORT SKIRT \\J THICKNE:

to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Figure 1-2. FNP1 &2 Pressurizer Surge Nozzle

.,..---.---~- ----------,

Z '3 39---------------~--'

t II 9

R Figure 1-3. FNP1 &2 Pressurizer Spray Nozzle "r----..----.----------,

Figure 1-4. FNP1 &2 Pressurizer SRV Nozzle

"..-~-----.--------,

z..

'1'--------------'~

i R References 1-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

1-2.

SI Calculation No. 2200205.301, "Transient Cycle Counts for Vogtle & Farley Plants,"

Revision 0.

Al-9

ATTACHMENT 2 PLANT-SPECIFIC APPLICABILITY VEGP UNITS 1 AND 2 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Section 9 of Reference [2-1] 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 VEGP1 &2 is provided in Table 2-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report are applicable to VEGP1 &2.

Table 2-1 Applicability of Reference [2-1] Representative Analyses to VEGP1&2 Pressurizer Surge Nozzle and Bottom Head Welds (Item Nos. B2.11 1 B2.12 and B3.110)

Category Requirement from Reference [2-1]

Applicability to VEGP1 &2 General The plant-specific pressurizer general transients As shown in Table 2-3, the VEGP1 &2 Requirements and cycles must be bounded by those shown in general transients are bounded by the Table 5-6 for a 60-year operating life. It should transients listed in Table 5-6 of be noted that the number of cycles were Reference (2-1].

extrapolated to 80 years in the evaluations.

The materials of the pressurizer shell and surge The VEGP1 &2 pressurizer shell is nozzle must be low alloy ferritic steels which fabricated from SA-533 Gr A Cl 2 conform to the requirements of ASME Code, material and the surge nozzle isSection XI, Appendix G, Paragraph G-2110.

fabricated from SA-508 Cl 2a material.

The associated RTNoT values are 60°F, which is bounded by the values used in Reference (2-1].

Both materials are low alloy ferritic steels that conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific pressurizer surge nozzle and The VEGP1 &2 weld configurations are Requirements bottom head weld configurations must conform shown in Figures 2-1, 2-3 and 2-4 and to those shown in Figure 1-1 (Item No. B2.11 ),

show conformance with the Figures Figure 1-2 (Item No. B2.12) and Figures 1-4, 1-shown in Reference (2-1 ].

5 or 1-6 (Item No. B3.110) of Reference (2-1 ].

The plant-specific dimensions of the pressurizer As shown in Table 2-2, the VEGP1 &2 shell and the surge nozzle must be within the pressurizer shell and surge nozzle range of values listed in Table 9-1 of Reference dimensions are within the range of (2-1 ].

values listed in Table 9-1 of Reference (2-1 ].

A2-1 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Category Requirement from Reference [2-1]

The plant-specific lnsurge/Outsurge transient definitions (temperature difference between the pressurizer shell and the pressurizer surge nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference (2-1].

Applicability to VEGP1 &2 As shown in Table 2-4, the VEGP1 &2 lnsurge/Outsurge transients are bounded by the transients listed in Table 5-10 of Reference (2-1].

Pressurizer Top Head Welds

{Item Nos. B2.11, B2.12 and B3.110)

Category Requirement from Reference [2-1]

Applicability to VEGP1 &2 General The plant-specific pressurizer general transients As shown in Table 2-3, the VEGP1 &2 Requirements and cycles must be bounded by those shown in general transients are bounded by the Table 5-6 of Reference (2-1] for a 60-year transients listed in Table 5-6 of operating life. It should be noted that the Reference (2-1].

number of cycles were extrapolated to 80 years in the evaluations.

The materials of the pressurizer shell and top The VEGP1 &2 pressurizer shell is head nozzles must be low alloy ferritic steels fabricated from SA-533 Gr A Cl 2 which conform to the requirements of ASME material and the top head nozzles are Code,Section XI, Appendix G, Paragraph G-fabricated from SA-508 Cl 2a. The 2110.

associated RT NOT values are 60°F, which is bounded by the values used in Reference (2-1].

Both materials are low alloy ferritic steels that conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific pressurizer top head nozzle The VEGP1 &2 top head nozzle weld Requirements weld configurations must conform to those configurations are shown in Figures 2-shown in Figure 1-1 (Item No. B2.11), Figure 1-1, 2-2, 2-5 and 2-6 and conform to the 2 (Item No. B2.12), Figure 1-3 (Item Nos. B2.21 Figures shown in Reference (2-1 ].

and B2.22) and Figures 1-4, 1-5 or 1-6 (Item No. B3.110) of Reference (2-1].

A2-2 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Category Requirement from Reference [2-1]

Applicability to VEGP1 &2 The plant-specific dimensions of the pressurizer As shown in Table 2-2, the VEGP1 &2 shell and the top head nozzles must be within pressurizer shell and top head nozzle the range of values listed in Table 9-1 of dimensions are within the range of Reference [2-1].

values listed in Table 9-1 of Reference

[2-1 ].

Table 2-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with VEGP1 &2 Component Geometric Parameter For a Westinghouse Plant VEGP1&2 Dimensions Pressurizer Inside Diameter (in)

Must be between 80 and 88 84.31 "

Shell NPS of piping or component (e.g.,

Surge Nozzle reducer) attached to nozzle safe-end Must be between 12 and 18 14" (in) (1)

Safety/Relief NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end Must be between 4 and 8 6"

Nozzle (in) (1)

NPS of piping or component (e.g.,

Spray Nozzle reducer) attached to nozzle safe-end Must be between 4 and 6 4"

(in) (1)

Note:

(1) Depending on the plant-specific configuration, the NPS of the piping (or component) attached directly to the nozzle (i.e., no safe-end) or to a safe-end extension may be used.

A2-3 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Table 2-3 Comparison of VEGP1&2 General Transients to Requirements in Reference [2-1]

Number of Cycles for 60 VEGP1 VEGP2 Transient Years from 60-Year 60-Year Table 5-6 of Projection Projection Reference [2-1]

Heatup / Cooldown 300 I 300 67 I 64 73 I 71 Loss of Load (Large Step Load Decrease, Loss of 360 103(1) 97(1)

Power, Loss of Flow, Reactor Trip)

Notes:

(1) Loss of Load transient bundled to conservatively envelope a combination of the following transients:

Loss of Load w/o Rx Trip Loss of RC Flow 1 Loop @ Power Large Step Load Decrease Reactor Trip (CD and SI)

Reactor Trip (CD no SI)

Reactor Trip (No Cooldown)

Table 2-4 Comparison of VEGP1&2 lnsurge/Outsurge Transients to Reference [2-1] Requirements 60-Year No. of Cycles From Table 5-10 of VEGP1 VEGP2 LH (°F)(1)

Reference [2-1] (For 60-Year 60-Year Westinghouse and CE Projection(2l Projection(2l Plants) 330 600 6

0 320 3,000 44 45 103 1,500 87 159 Notes:

(1) b.T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

(2) Table 15 of Reference [2-2].

A2-4 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Table 2-5 VEGP1 Inspection History Item Component No.

ID B2.11 11201-V6-002-W01 B2.11 11201-V6-002-W01 B2.11 11201-V6-002-W05 B2.11 11201-V6-002-W05 B2.12 11201-V6-002-W06 B2.12 11201-V6-002-W06 B2.12 11201-V6-002-W09 B2.12 11201-V6-002-W09 B3.110 11201-V6-002-W10 B3.110 11201-V6-002-W10 B3.110 11201-V6-002-W11 B3.110 11201-V6-002-W11 B3.110 11201-V6-002-W12 B3.110 11201-V6-002-W12 B3.110 11201-V6-002-W13 B3.110 11201-V6-002-W13 B3.110 11201-V6-002-W14 B3.110 11201-V6-002-W14 B3.110 11201-V6-002-W16 B3.110 11201-V6-002-W16 A= Acceptable Notes:

Exam Date 9/25/2018 9/26/2009 3/9/2023 9/23/2015 9/25/2018 9/26/2009 9/23/2015 3/24/2005 9/20/2018 9/24/2009 9/20/2018 9/24/2009 9/17/2021 3/11/2011 9/17/2021 3/11/2011 9/24/2015 3/21/2005 9/30/2015 3/25/2005 Interval/Period/

Exam Outage Results 4th//1 st/1 R21 A

3rd/1sU1R15 A

4th/2nd/1 R24 A

3rd/3rd/1 R 19 A(1l 4th/1 st/1 R21 A

3rd/1sU1R15 A

3rd/3rd/1 R 19 A

2nd/3rd/1 R 12 A

4th/1 st/1 R21 A

3rd/1sU1R15 A

4th/1 st/1 R21 A

3rd/1sU1R15 A

4th/2nd/1 R23 A

3rd/2nd/1 R 16 A

4th/2nd/1 R23 A

3rd/2nd/1 R 16 A

3rd/3rd/1 R 19 A

2nd/3rd/1 R 12 A

3rd/3rd/1 R 19 A

2nd/3rd/1 R 12 A

Coverage Relief Request 79.5%

Yes 75%

Yes 93.3%

n/a 88%

Yes, RR-3

> 90%

n/a

> 90%

n/a

> 90%

n/a 100%

n/a 55%

Yes 51.8%

Yes 55%

Yes 51.8%

Yes 64.51%

Yes 55.9%

Yes 65.22%

Yes 55.9%

Yes 48%

Yes 50%

Yes, RR-7 44.6%

RR-3 15%

Yes (1) Indications compared to previous data from 2005 and showed no signs of growth. Spot indications had no through-wall dimensions. Determined to be acceptable.

A2-5 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Item Component No.

ID B2.11 21201-V6-002-W01 B2.11 21201-V6-002-W01 B2.11 21201-V6-002-W01 B2.11 21201-V6-002-W05 B2.11 21201-V6-002-W05 B2.11 21201-V6-002-W05 B2.12 20201-V6-002-W06 B2.12 20201-V6-002-W06 B2.12 20201-V6-002-W09 B2.12 20201-V6-002-W09 B3.110 20201-V6-002-W10 B3.110 20201-V6-002-W10 B3.110 20201-V6-002-W11 B3.110 20201-V6-002-W11 B3.110 20201-V6-002-W12 B3.110 20201-V6-002-W12 B3.110 20201-V6-002-W13 B3.110 20201-V6-002-W13 B3.110 20201-V6-002-W14 B3.110 20201-V6-002-W14 B3.110 20201-V6-002-W16 B3.110 20201-V6-002-W16 A= Acceptable Notes:

Table 2-6 VEGP2 Inspection History Exam Interval/Period/

Date Outage 3/11/2022 4th/2nd/2R22 3/13/2019 4th/1 st/2R20 3/12/2010 3rd/1sU2R14 3/17/2019 4th/1 st/2R20 3/10/2016 3rd/3rd/2R 18 9/24/2005 2nd/3rd/2R 11 3/13/2019 4th/1 st/2R20 3/12/2010 3rd/1sU2R14 3/10/2016 3rd/3rd/2R 18 9/24/2005 2nd/3rd/2R 11 3/12/2019 4th/1 st/2R20 3/11/2010 3rd/1sU2R14 3/12/2019 4th/1 st/2R20 3/11/2010 3rd/1sU2R14 3/14/2022 4th/2nd/2R22 9/22/2011 3rd/2nd/2R 15 3/10/2022 4th/2nd/2R22 9/22/2011 3rd/2nd/2R 15 9/22/2011 3rd/2nd/2R 15 10/17/2002 2nd/2nd/2R9 3/13/2010 3rd/ 1 sU2VR 14 4/18/2001 2nd/2nd/2R8 Exam Coverage Results A

> 90%

A(1)

> 90%

A 75%

A 84%

A 84%

A 84%

A

> 90%

A

> 90%

A

> 90%

A

> 90%

A 49.5%

A 51.8%

A 49.5%

A 51.8%

A 65.22%

A 55.9%

A 65.22%

A 50%

A 55.9%

A 50%

A 15%

A 15%

(1) Laminar indication determined to be acceptable by Table IWB-3510-2.

A2-6 Relief Request n/a n/a

Yes, RR-7 Yes Yes Yes n/a n/a n/a n/a Yes Yes Yes RR-3 Yes RR-3 Yes Yes Yes Yes Yes Yes to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Figure 2-1. VEGP1 &2 Pressurizer Vessel A2-7 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) 11TrP g,

.38 MUl uETAIL *14 UEPER Wi~ TO UPPER[ WELD Figure 2-2. VEGP1 &2 Item No. B2.10 Weld Configuration - Upper Shell/Head CLA001Hli DETAIL ~c LOWER tEAD TO UMER st;£LL WELD Figure 2-3. VEGP1 &2 Item No. B2.10 Weld Configuration - Lower Shell/Head A2-8 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Figure 2-4. VEGP1 &2 Pressurizer Surge Nozzle l'l\\tSUUl[ft COOPUNll 1-P7:

Figure 2-5. VEGP1 &2 Pressurizer Spray Nozzle A2-9 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) ll.00

-'.5. _/-

cu l1l11WI....

Figure 2-6. VEGP1 &2 Pressurizer SRV Nozzle References 2-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

2-2.

SI Calculation No. 2200205.301, "Transient Cycle Counts for Vogtle & Farley Plants,"

Revision 0.

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ATTACHMENT 3 RESULTS OF INDUSTRY SURVEY to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Overall Industry Inspection Summary The results of an industry survey of past inspections of pressurizer welds are summarized in Reference [3-1]. Table 3-1 provides a summary of the combined survey results for Item Nos. B2.11, B2.12, B2.21, B2.22 and B3.110. The results identify that pressurizer examination of the items adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international 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 1,162 examinations for the components of the affected Item Nos. were conducted on PWR pressurizer components.

A small number of flaws were identified during these examinations which required flaw evaluation. None of these flaws were found to be service induced. Out of a total of 1,162 examinations identified by the plants that responded to the survey that have been performed on the above item numbers, only four examinations (for Item No. B2.11 ), at two units of a single plant site, identified flaws exceeding the acceptance criteria of ASME Code,Section XI. Flaw evaluations were performed to show acceptability of these indications and follow-on examinations showed no change in flaw sizes since the original inspections. No other indications were identified in any in-scope components.

A3-1 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Table 3-1 Summary of Survey Results Item No.

No. of Examinations No. of Reportable Indications B2.11 269 4 (1)

B2.12 269 0

B2.21 4

0 B2.22 30 0

B3.110 590 0

Note:

(1) Flaw evaluations were performed to show acceptability of these indications and follow-on examinations showed no change in flaw sizes since the original inspections. References 3-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

A3-2

ATTACHMENT 4 EVALUATION OF 15% MINIMUM COVERAGE FOR SNC PLANTS PRESSURIZER WELDS to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Background and Objective Some welds at the SNC plants (notably at VEGP1&2) have previous examination coverage as low as 15%. Since this is below the minimum coverage of 25% generically addressed in Reference [4-2], an evaluation is needed to address this minimum coverage. The objective of this attachment is to perform a sensitivity study on inspection coverage to determine the probabilities of rupture and leakage for the critical Case ID for the pressurizer welds (PRSHC-BW-2C) to address the minimum coverage of 15% for the SNC plants.

Technical Approach The technical approach follows that used in Reference [4-2] for the 25% minimum coverage evaluation. From Reference [4-2], the critical Case ID considered is PRSHC-BW-2C from Reference [4-1]. This same Case ID is evaluated for the 15% coverage for the SNC plants. The technical approach in Reference [4-2] is used for the evaluations together with observations made by the NRC in the Safety Evaluation (SE) [4-3] for the lead plant submittal [4-4]. The following values consistent with the technical basis documents [4-4-1] and the Reference [4-3]

SE will be used in the evaluation:

1. Weld Flaw Density - A value of 1.0 flaws per weld consistent with that recommended by the NRC in Section 3.8.9 of Reference [4-4-7].

2. Fracture Toughness - A mean value of 200 ksi in and a standard deviation 5 ksiin as recommended by the NRC in Section 6 of Reference [4-3].

3. Stress Multiplier - A stress multiplier of 1.0.

4. Inspection History - Plant specific inspection histories for FNP1 &2 and VEGP1 &2.

The PSI/ISi scenario considered for FNP1&2 is (PSl+10+20+30+40 70). The PSI/ISi scenario considered for VEGP1 &2 is therefore (PSl+10+20+30+60). In addition, the current ASME Code,Section XI PSI/ISi scenario of (PSl+10+20+30+40+50+60+70) will also be evaluated for comparison.

PROMISE Version 2.0 [4-5], which was used in the Reference [4-1] report and the Reference

[4-2] evaluations, will also be used in the sensitivity study on 15% minimum coverage for the limiting critical Case ID at the SNC plants.

Similar to the acceptance criteria in Reference [4-1], the probability of rupture and probability of leakage should equal to or less than 1.Ox10-5 in this evaluation.

Assumptions

1. 100% inspection coverage was achieved for preservice inspections (PSI) consistent with the assumption and ensuing discussions in Section 8.3.5 of Reference [4-1].

2. In-service inspections (ISi) at all plants are performed using ASME Code,Section XI procedures and therefore the POD curve used in the Reference [4-1] report is applicable.

3. The POD curve for ISi can be conservatively used for PSI as discussed in Attachment 1 of Reference [4-9] (pages A1-2 to A1-4 (Response to Open Audit Item 2.c.i.A)).

4. The critical Case ID (PRSHC-BW-2C) identified in the Reference [4-1] report and used in the Reference [4-2] evaluations will be used. Since this is the critical Case ID, all other Case IDs considered in Reference [4-1] will therefore be bounded.

A4-1 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1)

Results The results of the evaluation in terms of probability of rupture and probability of leakage are provided in Table 4-1. (Note: The selection of 80 years is for consistency with previous similar submittals. The number of years used in the analysis is not intended to involve time-limited assumptions defined by the current or future operating term due to the margin between the calculated probabilities and the acceptance criterion of 1.ox10-5 / year.)

Case ID PRSHC-BW-2C Table 4-2. Sensitivity Study for 15% ISi Coverage for FNP1 &2 and VEGP1 &2 Limiting Case PRSHC-BW-2C from Reference [4-1]

Probability of Probability of Rupture Leakage Scenario PSI/ISi Schedule (per Year) at 80 (per Year) at 80 years years 15% ISi 15% ISi Coverage Coverage VEGP1&2 PSl+10+20+30+60 1.250E-09 1.893E-06 FNP1&2 PSl+10+20+30+40+70 1.250E-09 1.891 E-06 ASME SXI PSl+10+20+30+40+50+60+70 1.250E-09 1.871 E-06 As shown in Table 4-1, the probabilities of rupture for the SNC plant-specific scenarios with ISi coverage of 15% is identical and remain unchanged from the current ASME Code,Section XI inspection scenario. The probability of rupture value of 1.25x10-9 for the two scenarios in Table 4-1 is approximately three orders of magnitude less than the acceptance criteria of 1.ox10-5 therefore ensuring adequate safety.

The probabilities of leakage for the SNC plant specific scenarios and the ASME Code,Section XI scenario assuming 15% ISi coverage are approximately 1.9x 10-5, which is above the acceptance criterion. As discussed in Sections 8.2.4.1.1 and 8.2.5 of Reference [4-6], the fact that the probability of leakage at a location slightly exceeds the acceptance criterion does not compromise plant safety. This is because pressure boundary leakage is detectable by plant operators, plant procedures allow for safe plant shutdown once any leakage is detected, and the probability of rupture values are maintained well below the acceptance criterion for 80 years of operation even under this limiting PSI/ISi scenario. This is consistent with the NRC staff acknowledgement in Section 3.8.7, page 19 of the Reference [4-8] based on the Reference [4-7] Request for Alternative, that even though the probability of leakage increased to a value greater than 1.Ox10-5 per year due to limited coverage less the 100%, leakage is not component rupture and would be managed by the plant leakage detection system.

Furthermore, as seen in Table 4-1, there is essentially no difference in the probabilities of rupture and leakage between the current ASME Code,Section XI schedule with 15% coverage and the SNC plant specific scenarios involving the 30-year examination deferral. This demonstrates that there is essentially no change in risk with respect to coverage in changing from the ASME Code,Section XI schedule to the 30-year examination deferral for SNC plants.

References 4-1.

Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shel/-

to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905, ADAMS Accession No. ML21021A271.

A4-2 to NL-24-0227 Proposed Alternative GEN-ISi-AL T-2024-03 in Accordance with 10 CFR 50.55a(z)(1) 4-2.

SI Calculation 2201462.301, "Justification for Minimum Coverage of 25% For a 30-Year ISi Deferral," Revision 0.

4-3.

Letter from J. G. Danna (USNRC) to E. Carr (PSEG Nuclear LLC), "Salem Generating Station Unit Nos. 1 and No. 2 - Authorization and Safety Evaluation for Alternate Request No. SC-14R-200 (EPID L-2020-LLR-0103)," dated June 10, 2021, ADAMS Accession No. ML21145A189.

4-4.

Letter from P. R. Duke, Jnr. (PSEG Nuclear LLC) to U.S.NRC, "Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Items Number B2.11 and B2.12," dated August 5, 2020, ADAMS Accession No. ML20218A587.

4-5.

Structural Integrity Associates Report DEV1806.402, PROMISE 2.0 Theory and User's Manual, Revision 1.

4-6.

Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590, ADAMS Accession No. ML19347B107.

4-7.

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

4-8.

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

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

4-9.

Letter No. NL-20-1312 from C. A. Gayheart (Southern Nuclear) to USNRC, "Vogtle Electric Generating Plant, Units 1 & 2, Response to Request for Additional Information Related to Proposed lnservice Inspection Alternative VEGP-ISI-AL T-04-04 Version 2.0,"

dated November 23, 2020, ADAMS Accession No. ML20329A302.

A4-3