NL-24-0201, Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)

Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)

ML24170B057
Person / Time
Site:	Farley, Vogtle
Issue date:	06/18/2024
From:	Coleman J Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-24-0201
Download: ML24170B057 (1)

Text

Regulatory Affairs 3535 Colonnade Parkway Birmingham AL 35243 205 992 5000 June 18, 2024 Docket Nos.: 50-348 50-424 NL-24-0201 50-364 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant, Units 1 and 2 Vogtle Electric Generating Plant, Units 1 and 2 Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator 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-ALT-2024-002 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 Category B-B and Table IWC-2500-1 Examination Category C-A for item numbers B2.40, C1.10, C1.20, and C1.30 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.

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-24-0201 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-AL T-2024-02 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 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 Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-1 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class:

Class 1 and Class 2

==

Description:==

Steam generator (SG) pressure-retaining welds Examination Category:

Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 2, Category C-A, pressure-retaining welds in pressure vessels Item Numbers:

B2.40 - Steam generators (primary side), tubesheet-to-head weld C1.10 - Shell circumferential welds C1.20 - Head circumferential welds C1.30 - Tubesheet-to-shell weld Farley Nuclear Plant Unit 1 (FNP1)

SG ASME Category ASME Item No.

Component ID Component Description A

B-B B2.40 ALA1-3100-1R Channel Head to Tubesheet A

C-A C1.10 ALA2-3100-4R Transition Cone to Lower Shell A

C-A C1.10 ALA2-3100-5R Upper Shell to Transition Cone A

C-A C1.20 ALA2-3100-6R Elliptical Head to Upper Shell A

C-A C1.30 ALA2-3100-2R Lower Shell Barrel to Upper Tubesheet B

B-B B2.40 ALA1-3200-1R Channel Head to Tubesheet B

C-A C1.10 ALA2-3200-4R Transition Cone to Lower Shell B

C-A C1.10 ALA2-3200-5R Upper Shell to Transition Cone B

C-A C1.20 ALA2-3200-6R Elliptical Head to Upper Shell B

C-A C1.30 ALA2-3200-2R Lower Shell Barrel to Upper Tubesheet C

B-B B2.40 ALA1-3300-1R Channel Head to Tubesheet C

C-A C1.10 ALA2-3300-4R Transition Cone to Lower Shell C

C-A C1.10 ALA2-3300-5R Upper Shell to Transition Cone C

C-A C1.20 ALA2-3300-6R Elliptical Head to Upper Shell C

C-A C1.30 ALA2-3300-2R Lower Shell Barrel to Upper Tubesheet Farley Nuclear Plant Unit 2 (FNP2)

SG ASME Category ASME Item No.

Component ID Component Description A

B-B B2.40 APR1-3100-1R Channel Head to Tubesheet A

C-A C1.10 APR2-3100-4R Transition Cone to Lower Shell

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-2 Farley Nuclear Plant Unit 2 (FNP2)

A C-A C1.10 APR2-3100-5R Upper Shell to Transition Cone A

C-A C1.20 APR2-3100-6R Elliptical Head to Upper Shell A

C-A C1.30 APR2-3100-2R Lower Shell Barrel to Upper Tubesheet B

B-B B2.40 APR1-3200-1R Channel Head to Tubesheet B

C-A C1.10 APR2-3200-4R Transition Cone to Lower Shell B

C-A C1.10 APR2-3200-5R Upper Shell to Transition Cone B

C-A C1.20 APR2-3200-6R Elliptical Head to Upper Shell B

C-A C1.30 APR2-3200-2R Lower Shell Barrel to Upper Tubesheet C

B-B B2.40 APR1-3300-1R Channel Head to Tubesheet C

C-A C1.10 APR2-3300-4R Transition Cone to Lower Shell C

C-A C1.10 APR2-3300-5R Upper Shell to Transition Cone C

C-A C1.20 APR2-3300-6R Elliptical Head to Upper Shell C

C-A C1.30 APR2-3300-2R Lower Shell Barrel to Upper Tubesheet Vogtle Electric Generating Plant Unit 1 (VEGP1)

SG ASME Category ASME Item No.

Component ID Component Description 1

B-B B2.40 11201-B6-001-W08 Tube Plate to Channel Head 1

C-A C1.10 11201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 1

C-A C1.10 11201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 1

C-A C1.20 11201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 1

C-A C1.30 11201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 2

B-B B2.40 11201-B6-002-W08 Tube Plate to Channel Head 2

C-A C1.10 11201-B6-002-W03 Upper Shell Barrel C to Transition Cone Weld 2

C-A C1.10 11201-B6-002-W04 Transition Cone to Lower Cone End Stub Barrel Weld 2

C-A C1.20 11201-B6-002-W01 Upper Head to Upper Shell Barrel D Weld 2

C-A C1.30 11201-B6-002-W07 Lower Shell Barrel A to Tube Plate Weld 3

B-B B2.40 11201-B6-003-W08 Tube Plate to Channel Head 3

C-A C1.10 11201-B6-003-W03 Upper Shell Barrel C to Transition Cone Weld 3

C-A C1.10 11201-B6-003-W04 Transition Cone to Lower Cone End Stub Barrel Weld 3

C-A C1.20 11201-B6-003-W01 Upper Head to Upper Shell Barrel D Weld

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-3 Vogtle Electric Generating Plant Unit 1 (VEGP1) 3 C-A C1.30 11201-B6-003-W07 Lower Shell Barrel A to Tube Plate Weld 4

B-B B2.40 11201-B6-004-W08 Tube Plate to Channel Head 4

C-A C1.10 11201-B6-004-W03 Upper Shell Barrel C to Transition Cone Weld 4

C-A C1.10 11201-B6-004-W04 Transition Cone to Lower Cone End Stub Barrel Weld 4

C-A C1.20 11201-B6-004-W01 Upper Head to Upper Shell Barrel D Weld 4

C-A C1.30 11201-B6-004-W07 Lower Shell Barrel A to Tube Plate Weld Vogtle Electric Generating Plant Unit 2 (VEGP2)

SG ASME Category ASME Item No.

Component ID Component Description 1

B-B B2.40 21201-B6-001-W08 Tube Plate to Channel Head 1

C-A C1.10 21201-B6-001-W03 Upper Shell Barrel C to Transition Cone Weld 1

C-A C1.10 21201-B6-001-W04 Transition Cone to Lower Cone End Stub Barrel Weld 1

C-A C1.20 21201-B6-001-W01 Upper Head to Upper Shell Barrel D Weld 1

C-A C1.30 21201-B6-001-W07 Lower Shell Barrel A to Tube Plate Weld 2

B-B B2.40 21201-B6-002-W08 Tube Plate to Channel Head 2

C-A C1.10 21201-B6-002-W03 Upper Shell Barrel C to Transition Cone Weld 2

C-A C1.10 21201-B6-002-W04 Transition Cone to Lower Cone End Stub Barrel Weld 2

C-A C1.20 21201-B6-002-W01 Upper Head to Upper Shell Barrel D Weld 2

C-A C1.30 21201-B6-002-W07 Lower Shell Barrel A to Tube Plate Weld 3

B-B B2.40 21201-B6-003-W08 Tube Plate to Channel Head 3

C-A C1.10 21201-B6-003-W03 Upper Shell Barrel C to Transition Cone Weld 3

C-A C1.10 21201-B6-003-W04 Transition Cone to Lower Cone End Stub Barrel Weld 3

C-A C1.20 21201-B6-003-W01 Upper Head to Upper Shell Barrel D Weld 3

C-A C1.30 21201-B6-003-W07 Lower Shell Barrel A to Tube Plate Weld 4

B-B B2.40 21201-B6-004-W08 Tube Plate to Channel Head 4

C-A C1.10 21201-B6-004-W03 Upper Shell Barrel C to Transition Cone Weld 4

C-A C1.10 21201-B6-004-W04 Transition Cone to Lower Cone End Stub Barrel Weld 4

C-A C1.20 21201-B6-004-W01 Upper Head to Upper Shell Barrel D Weld 4

C-A C1.30 21201-B6-004-W07 Lower Shell Barrel A to Tube Plate Weld

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-4 2.0 APPLICABLE CODE EDITION AND ADDENDA:

FNP Units 1 and 2 The fifth 10-year inservice inspection (ISI) interval Code of Record for FNP Units 1 and 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 Inservice Inspection of Nuclear Power Plant Components. The current fifth 10-year interval start date was December 1, 2017 and will end November 30, 2027.

VEGP Units 1 and 2 The fourth ISI interval Code of Record for VEGP Units 1 and 2 is the 2007 Edition of ASME Boiler and Pressure Vessel Code,Section XI, with 2008 Addenda, Rules for Inservice 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:

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

Item No. B2.40 - Volumetric examination of essentially 100% of the weld length of all welds during the first Section XI inspection interval. For successive inspection intervals the examination may be limited to one vessel among the group of vessels performing a similar function. The examination volume is shown in Figure IWB-2500-6.

Item No. C1.10 - Volumetric examination of essentially 100% of the weld length of the cylindrical-shell-to-conical shell-junction welds and shell (or head)-to-flange welds 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 Figure IWC-2500-1.

Item No. C1.20 - Volumetric examination of essentially 100% of the weld length of the head-to-shell weld 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 Figure IWC-2500-1.

Item No. C1.30 - Volumetric examination of essentially 100% of the weld length of the tubesheet-to-shell welds 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 Figure IWC-2500-2.

4.0 REASON FOR REQUEST:

The Electric Power Research Institute (EPRI) performed an assessment in Reference

[9.1] of the bases for the ASME Code,Section XI examination requirements specified for

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-5 the above listed ASME Code,Section XI, Division 1 examination categories for SG welds and components. The assessment includes 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 deferred for some time with no impact to plant safety. Based on the conclusions of the EPRI report, supplemented by the plant-specific evaluations contained In Attachments 1 and 2, Southern Nuclear Company (SNC) is requesting an ISI examination deferral for the subject welds. The Reference [9.1] report was developed consistent with the recommendations provided in EPRIs White Paper on suggested content for PFM submittals (Reference [9.12]) and NRC Regulatory Guide 1.245 for PFM submittals and associated technical basis (References [9.13 and 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 Code,Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.10 Shell circumferential welds C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld In 2000 (first period of the third inspection interval) the FNP1 SGs were replaced. The replacement SG (RSG) welds and components received the required preservice inspection (PSI) examinations prior to service followed by ISI examinations through the first period of the current fifth inspection interval. It should be noted that during the fourth inspection interval, the Code of Record was the 2001 Edition of Section XI through the 2003 Addenda. This Code of Record only required Category C-A, Item C1.10 exams to be performed on shell circumferential welds at a gross structural discontinuity. The RSGs do not have a shell circumferential weld at a gross structural discontinuity, and therefore no Category C-A, Item C1.10 exams were performed during the fourth inspection interval.

The proposed alternative is to defer the ISI examinations for the above listed Item Nos.

for the FNP1 RSGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fifth ISI interval and to inspections scheduled for the sixth ISI interval. The sixth ISI interval is currently scheduled to end on November 30, 2037. All examinations will be scheduled to occur in the same interval period as the last examination, but with a two-interval inspection periodicity.

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-6 FNP2 For FNP2, SNC is requesting an inspection alternative to the examination requirements of ASME Code,Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.10 Shell circumferential welds C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld In 2001 (third period of the second inspection interval) the FNP2 SGs were replaced.

The RSG 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. Note that in 2008, Unit 2 interval dates were synchronized with Unit 1 interval dates, as described in Reference [9.31]. It should also be noted that during the fourth inspection interval, the Code of Record was the 2001 Edition of Section XI through the 2003 Addenda. This Code of Record only required Category C-A, Item C1.10 exams to be performed on shell circumferential welds at a gross structural discontinuity. The RSGs do not have a shell circumferential weld at a gross structural discontinuity, and therefore no Category C-A, Item C1.10 exams were performed during the fourth inspection interval.

The proposed alternative is to defer the ISI examinations for the above listed Item Nos.

for the FNP2 RSGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fifth ISI interval and to inspections scheduled for the sixth ISI interval. The sixth ISI interval is currently scheduled to end on November 30, 2037. All examinations will be scheduled to occur in the same interval 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 Code,Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.10 Shell circumferential welds C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-7 VEGP1 still has its original SGs. The SG 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 the above listed Item Nos.

for the VEGP1 SGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fourth 10-year ISI interval and to inspections scheduled during the fifth ISI interval. The fifth ISI interval is currently scheduled to end on May 30, 2037. All examinations will be scheduled to occur in the same interval 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 Code,Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.10 Shell circumferential welds C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld VEGP2 still has its original SGs. The SG 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 the above listed Item Nos.

for the VEGP2 SGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fourth 10-year ISI interval and to inspections scheduled during the fifth ISI interval. The fifth ISI interval is currently scheduled to end on May 30, 2037. All examinations will be scheduled to occur in the same interval 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 provided below.

The applicability of the technical basis to FNP Units 1 and 2 and VEGP Units 1 and 2 is shown in Attachments 1 and 2, respectively.

Applicability of the Degradation Mechanism Evaluation in Reference [9.1] to FNP Units 1 and 2 and VEGP Units 1 and 2 An evaluation of degradation mechanisms that could potentially impact the reliability of the SG welds and components was performed in Reference [9.1]. The degradation

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-8 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 known active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG welds and components covered in this request.

This observation was acknowledged by the NRC in Section 3.8, page 6, second paragraph of the Reference [9.16] Safety Evaluation (SE) for VEGP Units 1 and 2 and Section 2.0, page 3, second paragraph of the Reference [9.18] SE for Millstone Unit 2.

As shown in Attachments 1 and 2, the materials and operating conditions for FNP Units 1 and 2 and VEGP Units 1 and 2 are similar to those in the Reference [9.1]. Therefore, the conclusions of this Report apply to FNP and VEGP. 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 74 PWR 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 FNP Units 1 and 2 and VEGP Units 1 and 2 Finite element analyses (FEA) were performed in Reference [9.1] to determine the stresses in the SG welds and components covered in this request. The finite element models used in Reference [9.1] are consistent with the configurations at FNP Units 1 and 2 and VEGP Units 1 and 2 and therefore no new FEA model is required for the stress analysis of these plants. The analysis in Reference [9.1] 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 FNP Units 1 and 2 and VEGP Units 1 and 2 is demonstrated in Attachments 1 and 2 and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [9.1] stress analyses are compared to those of FNP Units 1 and 2 and VEGP Units 1 and 2 in Table 1:

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-9 Table 1: SG Vessel Dimensions Plant Primary Lower Head ID (in)

Primary Lower Head Thk (in)

Primary Lower Head Ri/t Secondary Upper Shell ID (in)

Secondary Upper Shell Thk (in)

Secondary Upper Shell Ri/t EPRI Report (Table 4-2 of Reference [9.1])

155.33 6.94 11.2 230.87 4.91 23.5 FNP1 125.56 5.22 12.03 168.5 4.21 20.0 FNP2 125.56 5.22 12.03 168.5 4.21 20.0 VEGP1 125.88 5.56 11.32 168.88 3.52 23.99 VEGP2 125.88 5.56 11.32 168.88 3.52 23.99 As discussed in Sections 4.3.3 and 4.6 of Reference [9.1] and noted by the NRC in Section 3.8.3.1, page 9, third paragraph of the SER for Vogtle [9.16], the dominant stress is the pressure stress. Therefore, the variation in the Ri/t ratio determined in Table 1 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. From Table 1, the stress ratio (Ri/t) of FNP Units 1 and 2 and VEGP Units 1 and 2 relative to the that used in the EPRI report are as follows:

FNP Units 1 and 2 primary lower head: (12.03/11.2) = 1.07 FNP Units 1 and 2 secondary upper shell: (20/23.5) = 0.85 VEGP Units 1 and 2 primary lower head: (11.32/11.2) = 1.01 VEGP Units 1 and 2 secondary upper shell: (23.99/23.5) = 1.02 These stress ratios are bounded by the 2.1 safety factor used in Reference 9.1.

Therefore, the stress analysis in Reference 9.1 is applicable to FNP Units 1 and 2 and VEGP Units 1 and 2.

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 FNP Units 1 and 2 and VEGP Units 1 and 2 are performed at normal operating conditions. No hydrostatic testing had been performed at FNP Units 1 and 2 and VEGP Units 1 and 2 since the SGs went into service.

In Reference [9.1], clad residual stress was not considered for the primary side welds.

In a previous NRC request for additional information (RAI) (Reference [9.19], RAI 3c),

the NRC raised this issue. In response to the RAI (Reference [9.20], RAI Response 3.c), an evaluation was performed which showed that the clad residual stress has no significant impact on the conclusions of Reference [9.1] and this was found acceptable by the NRC in Section 5.3 of Reference [9.18]

Applicability of the Flaw Tolerance Evaluation in Reference [9.1] to FNP Units 1 and 2 and VEGP Units 1 and 2

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-10 Flaw tolerance evaluations were performed in Reference [9.1] consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a PSI followed by subsequent ISIs, the NRCs safety goal of 1x10-6 failures per year is met.

The PFM analysis in Reference [9.2] 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 previous alternative request (for the SG FW and MS nozzle components covered by Reference [9.2]), 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 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 PSI/ISI scenarios for FNP Units 1 and 2 and VEGP Units 1 and 2 are discussed below. Note that the assumption below of a 30-year ISI deferral is conservative compared to the every other inspection interval being requested.

FNP1 For the FNP1 RSGs (replaced in the first period of the third ISI interval), PSI examinations have been performed followed by ISI examinations over two 10-year intervals as required by ASME Section XI (the unit is currently in its fifth ISI interval). The PSI/ISI scenario considered is therefore (PSI+10+20+50).

FNP2 For the FNP2 RSGs (replaced in the third period of the second ISI interval), PSI examinations have been performed followed by ISI examinations over two 10-year intervals as required by ASME Section XI (the unit is currently in its fifth ISI interval). The PSI/ISI scenario considered is therefore (PSI+10+20+50).

VEGP1 For the VEGP1 original SGs, PSI examinations have been performed followed by ISI examinations over three 10-year intervals as required by ASME Section XI (the unit is currently in its fourth ISI interval). The PSI/ISI scenario considered is therefore (PSI+10+20+30+60).

VEGP2 For the VEGP2 original SGs, PSI examinations have been performed followed by ISI examinations over three 10-year intervals as required by ASME Section XI (the unit is currently in its fourth ISI interval). The PSI/ISI scenario considered is therefore (PSI+10+20+30+60).

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-11 Limiting PSI/ISI Scenario The most limiting PSI/ISI scenario for FNP Units 1 and 2 and VEGP Units 1 and 2 is (PSI+10+20+50) since it has fewer inspections. This limiting PSI/ISI scenarios for FNP Units 1 and 2 and VEGP1&2 were 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, the following additional plant-specific evaluations were performed with the limiting PSI/ISI scenario shown for FNP Units 1 and 2 and VEGP1&2. These evaluations are described in Attachments 1 and 2.

For the SG welds being considered, Table 8-32 of Reference [9.1] indicates that the critical Case ID is SGPTH-P4A. This case was evaluated for the limiting inspection scenario of PSI+10+20+50, a flaw density of 1.0 flaw per weld, a fracture toughness of 200 ksiin and a standard deviation 5 ksiin. A relatively high stress multiplier of 2.1 was applied. The results of the evaluation, using PROMISE Version 2.0, are summarized in Table 2 and show that after 80 years, the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-6.

Table 2: Sensitivity to Combined Effects of Fracture Toughness, Stress, and Weld Flaw Density for 80 Years for the Remaining SG Welds (Westinghouse, CE or B&W)

(Case ID SGPTH-P4A from Reference [9.1])

Time (year)

Probability per Year for Combined Case KIC = 200 ksiin.

SD = 5 ksiin.

Stress Multiplier = 2.1 Weld Flaw Density = 1 PSI+10+20+50 Rupture Leak 10 2.20E-07 1.00E-08 20 2.20E-07 5.00E-09 30 1.53E-07 3.33E-09 40 1.43E-07 2.50E-09 50 3.24E-07 2.00E-09 60 2.70E-07 1.67E-09 70 2.40E-07 1.43E-09 80 2.26E-07 1.25E-09 The plant-specific PFM evaluation presented above for FNP Units 1 and 2 and VEGP Units 1 and 2 indicates that with conservative inputs of the critical parameters, the probabilities of rupture and leakage are well below the acceptance criterion of 1x10-6 failures per year. The stress multiplier of 2.1 applied in Table 2 is greater than the plant specific stress ratios of 1.07 for FNP Units 1 and 2 and 1.01 for VEGP Units 1 and 2 for the primary head and 0.85 and 1.02, respectively, for the secondary upper shell determined previously from the geometrical data in Table 1. Therefore, the stresses and fracture mechanics evaluations in the Reference [9.1] EPRI report are conservative in

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-12 application to FNP Units 1 and 2 and VEGP Units 1 and 2. It should also be noted that the evaluation incorporates conservative assumptions with regard to the PSI/ISI scenarios. Furthermore, the evaluation was performed for 80 years from the last inspection of the component, which is longer than the deferral being sought by SNC in this Request for Alternative.

An evaluation was performed to show acceptability of the low KIC values at the beginning and ending of the heatup/cooldown transient for the FW and MS nozzles to address Item No. 2.e.iii during the NRC audit of PROMISE [9.25]. The evaluation was performed using an RTNDT value of 60oF, the maximum allowed by BTP 5-3 [9.26]. The RTNDT value of 60oF bounds the value of 10 oF in Attachment 2 and 60oF in Attachment 3 for the SG materials. The evaluation showed acceptable results for the limiting Case IDs from the Reference [9.1] EPRI report. This was found acceptable by the NRC [9.27]. A similar evaluation was performed for the remainder of the SG welds in Reference [9.28] using the limiting Case ID from the Reference [9.2] EPRI report to address NRC RAI-6 in Reference [9.29]. In this evaluation, the limiting RTNDT value of 60oF was used and acceptable results were also obtained.

The PFM evaluations documented in Reference [9.1] and the plant-specific evaluations above used a Section XI, Appendix VIII-based probability of detection (POD) curve in the PFM evaluation because most ISI examinations of major plant Class 1 and Class 2 components are performed using Appendix VIII procedures. In the case of FNP Units 1 and 2 and VEGP Units 1 and 2, ASME Code,Section V procedures are used. Based on the observations made by the NRC in Section 3.8.8.2, page 21 of the Vogtle SE [9.16],

the use of the ASME Code,Section XI, Appendix VIII based POD curve for inspections based on ASME Code,Section V procedures would have minimal impact of the PFM results since the POD curve is not one of the parameters that significantly affect the PFM results.

The DFM evaluations in Table 8-3 of Reference [9.1] and Table 8-31 of Reference [9.2]

provide verification of the above PFM results for FNP Units 1 and 2 and VEGP Units 1 and 2 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] and Section 8.2.4.1.1 of Reference

[9.2], PSI refers to the collective examinations required by ASME Code,Section III during fabrication and any ASME Code,Section XI examinations performed prior to service. The Section III 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, all welds/components have examinations coverage greater

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-13 than 90% (essentially 100%). As shown in Attachment 1, no flaws that 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, all welds/components have examinations coverage greater than 90% (essentially 100%). 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, some welds/components have examinations coverage less than 90% but greater than 80%. An evaluation performed in Reference [9.30] indicated that inspection coverage as low as 25% is acceptable and therefore all inspection coverages for VEGP1 are acceptable. As shown in Attachment 1, 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, some welds/components have examinations coverage less than 90% but greater than 80%. An evaluation performed in Reference [9.30] indicated that inspection coverage as low as 25% is acceptable and therefore all inspection coverages for VEGP2 are acceptable. 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 (Reference [9.21]) which was found acceptable by the NRC (Reference [9.22]) 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

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-14 is in progress to adopt ASME Code Case N-921. Following the current inspection interval, both FNP Units 1 and 2 and VEGP Units 1 and 2 will update to at least the 2019 version of ASME 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 10 years. The below tables provide the schedule of future examinations with and without 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 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.

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-15 Previous and Next Scheduled Examination for Each Component FNP Units 1 and 2 Item No.

Date Interval

/ Period Components ID Next Exam-ination Interval

/

Period Next Examination Window without Code Case N-921 Next Examination Window with Code Case N-921 B2.40 1R30 4/2/2021 5th/2nd ALA1-3300-1R 7th/2nd 12/1/2040 -

11/30/2044 12/1/2043 -

11/30/2047 2R24 4/22/2016 4th/3rd APR1-3200-1R 6th/3rd 12/1/2034 -

11/30/2037 12/1/2035 -

11/30/2039 C1.10 1R29 10/11/2019 5th/1st ALA2-3100-4R 7th/1st 12/1/2037 -

11/30/2040 12/1/3039 -

11/30/2043 1R29 10/10/2019 5th/1st ALA2-3100-5R 7th/1st 12/1/2037 -

11/30/2040 12/1/3039 -

11/30/2043 2R29 10/20/2023 5th/2nd APR2-3100-4R 7th/2nd 12/1/2040 -

11/30/2044 12/1/2043 -

11/30/2047 2R29 10/20/2023 5th/2nd APR2-3100-5R 7th/2nd 12/1/2040 -

11/30/2044 12/1/2043 -

11/30/2047 C1.20 1R27 10/14/2016 4th/3rd ALA2-3300-6R 6th/3rd 12/1/2034 -

11/30/2037 12/1/2035 -

11/30/2039 2R26 4/18/2019 5th/1st APR2-3100-6R 7th/1st 12/1/2037 -

11/30/2040 12/1/3039 -

11/30/2043 C1.30 1R30 4/2/2021 5th/2nd ALA2-3300-2R 7th/2nd 12/1/2040 -

11/30/2044 12/1/2043 -

11/30/2047 2R25 10/23/2017 4th/3rd APR2-3100-2R 6th/3rd 12/1/2034 -

11/30/2037 12/1/2035 -

11/30/2039

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-16 Next Scheduled Examination for Each Component VEGP Units 1 and 2 Conclusion It is concluded that the SG pressure-retaining welds are very flaw tolerant. PFM and DFM evaluations performed as part of the Reference [9.1] technical basis report, supplemented by plant-specific evaluations performed as part of this Request for Item No.

Date Interval

/ Period Components ID Next Exam-ination Interval

/

Period Next Examination Window without Code Case N-921 Next Examination Window with Code Case N-921 B2.40 1R16 3/20/2011 3rd/2nd 11201-B6-001-W08 5th/2nd 5/31/2030 -

5/30/2034 5/31/2031 -

5/30/2035 2R17A 9/24/2014 3rd/2nd 21201-B6-004-W08 5th/2nd 5/31/2030 -

5/30/2034 5/31/2031 -

5/30/2035 C1.10 1R22 3/17/2020 4th/1st 11201-B6-002-W03 6th/1st 5/31/2037 -

5/30/2040 5/31/2039 -

5/30/2043 1R18 3/26/2014 3rd/3rd 11201-B6-002-W04 5th/3rd 5/31/3034 -

5/30/3037 5/31/2035 -

5/30/2039 2R17 9/26/2014 3rd/ 3rd 21201-B6-002-W04 5th/3rd 5/31/3034 -

5/30/3037 5/31/2035 -

5/30/2039 2R23 9/19/2023 4rd/2nd 21201-B6-002-W03 6th/2nd 5/31/2040 -

5/30/2044 5/31/2043 -

5/30/2047 C1.20 1R21 9/23/2018 4th/1st 11201-B6-001-W01 6th/1st 5/31/2037 -

5/30/2040 5/31/2039 -

5/30/2043 2R20 3/18/2019 4th/1st 21201-B6-001-W01 6th/1st 5/31/2037 -

5/30/2040 5/31/2039 -

5/30/2043 C1.30 1R24 3/7/2023 4th/2nd 11201-B6-001-W007 6th/2nd 5/31/2040 -

5/30/2044 5/31/2043 -

5/30/2047 2R23 9/16/2023 4th/2nd 21201-B6-004-W007 6th/2nd 5/31/2040 -

5/30/2044 5/31/2043 -

5/30/2047

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-17 Alternative, demonstrate that using conservative PSI/ISI inspection scenarios for all plants, the NRC safety goal of 10-6 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to FNP Units 1 and 2 and VEGP Units 1 and 2 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 inservice inspection interval 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 SG welds examined in the two most recent 10-year inspection intervals.

In addition to the required PSI examinations for these SG welds and components, FNP Units 1 and 2 and VEGP Units 1 and 2 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.

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 RSGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fifth ISI interval and to inspections scheduled for the sixth ISI interval. The sixth ISI interval is currently scheduled to end on November 30, 2037. All examinations will be scheduled to occur in the same interval period as the last examination, but with a two-interval inspection periodicity.

FNP2 The proposed alternative is to defer the ISI examinations for these Item Nos. for the FNP2 RSGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fifth ISI interval and to inspections scheduled for the sixth ISI interval. The sixth ISI interval is currently scheduled to end on November 30, 2037. All examinations will be scheduled to occur in the same interval period as the last examination, but with a two-interval inspection periodicity.

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-18 VEGP1 The proposed alternative is to defer the ISI examinations for these Item Nos. for the VEGP1 SGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fourth 10-year ISI interval and to inspections scheduled during the fifth ISI interval. The fifth ISI interval is currently scheduled to end on May 30, 2037. All examinations will be scheduled to occur in the same interval period as the last examination, but with a two-interval inspection periodicity.

VEGP2 The proposed alternative is to defer the ISI examinations for the above listed Item Nos.

for the VEGP2 SGs from the current ISI interval requirement to every other interval. This would apply to inspections required during the remainder of the fourth 10-year ISI interval and to inspections scheduled during the fifth ISI interval. The fifth ISI interval is currently scheduled to end on May 30, 2037. All examinations will be scheduled to occur in the same interval period as the last examination, but with a two-interval inspection periodicity.

7.0 PRECEDENTS

The following submittal was 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 [9.1] technical basis report:

Letter from C. A. Gayheart (SNC) to the 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 (Reference [9.15]).

The NRC issued a safety evaluation of the SNC request for alternative on January 11, 2021.

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 [9.16].

The following submittal was made by Dominion Energy to provide relief from the ASME Section XI Examination Category B-B (Item No. B2.40) and Category C-A (Item Nos.

C1.10, C1.20 and C1.30) surface and volumetric examinations based on the Reference

[9.2] technical basis report:

Letter from Mark D. Sartain (Dominion Energy) to the NRC, Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles, dated July 15, 2020, ADAMS Accession No. ML20198M682 (Reference [9.17]).

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-19 The NRC issued a safety evaluation of the Dominion Energy request for alternative on July 16, 2021.

Letter from James G. Danna (USNRC) to Daniel G. Stoddard (Dominion Energy),

Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-05-06 (EPID L-2020-LLR-0097), dated July 16, 2021, ADAMS Accession No. ML21167A355 [9.18].

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

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, dated July 24, 2000, ADAMS Accession No. ML003730922.

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

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, dated January 7, 2010, ADAMS Accession No. ML093561419.

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, dated March 12, 2012, ADAMS Accession No. ML120541062.

Letter from R. J. Pascarelli (NRC) to E. D. Halpin (PG&E), Diablo Canyon Plant, Units 1 and 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, dated December 8, 2015, ADAMS Accession No. ML15337A021.

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

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 (Reference [9.6]) and BWRVIP-241 (Reference [9.8]), 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 [9.7] and [9.9].

The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 (Reference [9.10]), which has been conditionally approved by the NRC in Revision 19 of Regulatory Guide 1.147 (Reference [9.11]).

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-20

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 POD Probability of detection PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor RSG Replacement Steam Generator SCC Stress corrosion cracking SG Steam Generator SNC Southern Nuclear Operating Company WEC Westinghouse Electric Company

9.0 REFERENCES

9.1 Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

9.2 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.

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, ADAMS Accession No. ML11304A110.

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-21 9.5 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 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 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.

9.10 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 NRC Regulatory Guide 1.147, Revision 18, Inservice 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 NRC Regulatory Guide 1.245, Revision 0, Preparing Probabilistic Fracture Mechanics Submittals, January 2022.

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

9.15 Letter from C. A. Gayheart (SNC) to the 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.

9.16 Letter from Michael T. Markley (NRC) to Cheryl A. Gayheart (SNC), 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, dated January 11, 2021, ADAMS Accession No. ML20352A155.

9.17 Letter from Mark D. Sartain (Dominion Energy) to the NRC, Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles, dated July 15, 2020, ADAMS Accession No. ML20198M682.

9.18 Letter from James G. Danna (NRC) to Daniel G. Stoddard (Dominion Energy),

Millstone Power Station Unit 2 - Authorization and Safety Evaluation for

Enclosure to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

E-22 Alternative Request No. RR-05-06, dated July 16, 2021, ADAMS Accession No. ML21167A355.

9.19 Not used.

9.20 Not used.

9.21 Letter from C. H. Gayheart (SNC) to NRC. Joseph M. Farley Nuclear Plant -

Units 1 and 2 Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0, dated September 30, 2022, ADAMS Accession No. ML22273A159.

9.22 Letter form J. M. Heisserer (NRC) to J. M. Coleman (SNC) Joseph M. Farley Nuclear Plant, Units 1 And 2 - Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0, to the Requirements of the ASME Code, dated August 30, 2023, ADAMS Accession No. ML23164A120.

9.23 Not used.

9.24 Not used.

9.25 Letter from J. G. Lamb (NRC) to C. A. Gayheart (SNC), Vogtle Electric Generating Plant, Units 1 and 2 - Audit Plan for Relief Request Inservice Inspection Alternative VEGP-ISI-ALT-04-04," dated May 14, 2020, ADAMS Accession No. ML20128J311.

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

9.27 Letter from J. G. Lamb (NRC) to C. A. Gayheart (SNC), Vogtle Electric Generating Plant, Units 1 and 2 - Audit Report for the PROMISE Version 1.0 Probabilistic Fracture Mechanics Software Code Used in Relief Request VEGP-ISI-ALT-04-04," dated December 10, 2020, ADAMS Accession No. ML20258A002.

9.28 Letter from D. T, Gudger (Constellation Energy Generation, LLC) to NRC, Response to Request for Additional Information - Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles, dated June 17, 2022, ADAMS Accession No. ML22168A005.

9.29 Email Letter from J. Wiebe (NRC) to T. Loomis (Constellation Energy Generation, LLC), Draft RAIs for Requests for Alternatives I4R-17, I4R-23, ISI-05-018, I6R-10, dated May 6, 2022, ADAMS Accession No. ML222129A013.

9.30 N. Cofie, D. Dedhia, S. Chesworth, D. J. Shim and R. Grizzi, Technical Basis for Inspection Optimization and Deferral of PWR Steam Generator Component Examinations, Paper No. PVP2023-105958, ASME Pressure Vessels and Piping Division Conference, July 16-21, 2023, Atlanta, Georgia, USA.

9.31 Letter from M. C. Wong (NRC) to J. R. Johnson (SNC), "Joseph M Farley Nuclear Plant, Unit 2, Safety Evaluation for Alternative FNP-ISI-ALT-01, Version 1.0 Relief Request from ASME Code Requirements," dated October 17, 2008, ADAMS Accession No. ML082940114.

ATTACHMENT 1 PLANT-SPECIFIC APPLICABILITY FNP UNITS 1 AND 2 to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A1-1 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 FNP Units 1 and 2 is provided in Table 1-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to FNP Units 1 and 2.

Table 1-1 Applicability of Reference [1-1] Representative Analyses to FNP1&2 Item No. B2.40 (SG Primary Side Welds)

Category Requirement from Reference [1-1]

Applicability to FNP1&2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of a portion of the vessel) is not considered in this evaluation due to its rarity.

In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance.

FNP1&2 have not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The FNP1&2 SG channel heads and tubesheet are fabricated from SA-508 Class 3 material. The associated RTNDT value is +10F, which is bounded by the value used in the EPRI Report.

SA-508 Class 3 is a low alloy ferritic steel which conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [1-1].

The FNP1&2 weld configurations are shown in Figure 1-1 and show conformance with Figure 1-2 of Reference [1-1].

The SG vessel dimensions must be within 10%

of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [1-1].

Per Table 1 in the main section of this request for alternative, the FNP1&2 SG vessel dimensions are as follows:

SG Lower Head OD = 136 to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A1-2 Category Requirement from Reference [1-1]

Applicability to FNP1&2 SG Upper Shell OD = 176.92 The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [1-1] for Westinghouse plants.

The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [1-1] over a 60-year operating life.

As shown in Table 1-2, the FNP1&2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-7 of Reference [1-1].

Items No. C1.10, C1.20 and C1.30 (SG Secondary Side Shell Welds)

Category Requirement from Reference [1-1]

Applicability to FNP1&2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of a portion of the vessel) is not considered in this evaluation due to its rarity.

In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance.

FNP1&2 have not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The FNP1&2 SG vessel shell and tubesheet are fabricated from SA-508 Class 3 material. The associated RTNDT value is +10F, which is bounded by the value used in the EPRI Report.

SA-508 Class 3 is a low alloy ferritic steel which conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [1-1].

The FNP1&2 weld configurations are shown in Figures 1-1 and conform to Figures 1-7 and 1-8 of Reference [1-1].

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A1-3 Category Requirement from Reference [1-1]

Applicability to FNP1&2 The SG vessel dimensions must be within 10%

of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [1-1].

Per Table 1 in the main section of this request for alternative, the FNP1&2 SG vessel dimensions are as follows:

SG Lower Head OD = 136 SG Upper Shell OD = 176.92 The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [1-1] for Westinghouse plants.

The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [1-1] over a 60-year operating life.

As shown in Tables 1-2 and 1-3, the FNP1&2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Tables 5-7 and 5-9 of Reference [1-1].

Table 1-2 FNP1&2 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Welds (Comparison to Table 5-7 of Reference [1-1])

Transient(1)

Number of Cycles for 60 Years from Table 5-7 of Reference [1-1]

FNP1 60-Year Projection(2)

FNP2 60-Year Projection(3)

Heatup/Cooldown 300 80/79 64/63 Plant Loading/Unloading 5000 318/335 278/483 Reactor Trip 360 98 96 Notes:

1.

Table 5-7 of Reference [1-1] also includes allowable transient temperatures and pressures. From previous experience with Westinghouse plants, these values are with 2% of the values used in the EPRI report. This is acceptable based on the large margins in the evaluation as demonstrated by the low plant specific stress ratio compared to the maximum allowed stress ratio.

2.

From Table 10 of Reference [1-2].

3.

From Table 11 of Reference [1-2].

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A1-4 Table 1-3 FNP1&2 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Welds (Comparison to Table 5-9 of Reference [1-1])

Transient(1)

Number of Cycles for 60 Years from Table 5-9 of Reference [1-1]

FNP1 60-Year Projection(2)

FNP2 60-Year Projection(3)

Heatup/Cooldown 300 80/79 64/63 Plant Loading/Unloading 5000 318/335 278/483 Reactor Trip 360 98 96 Notes:

1.

Table 5-9 of Reference [1-1] also includes allowable transient temperatures and pressures. From previous experience with Westinghouse plants, these values are with 2% of the values used in the EPRI report. This is acceptable based on the large margins in the evaluation as demonstrated by the low plant specific stress ratio compared to the maximum allowed stress ratio.

2.

From Table 10 of Reference [1-2].

3.

From Table 11 of Reference [1-2].

Table 1-4. FNP1 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.40 ALA1-3300-1R 4/12/2012 4th/2nd/1R24 A

> 90%

no B2.40 ALA1-3300-1R 4/2/2021 5th/2nd/1FR30 A

> 90%

no C1.10 ALA2-3100-4R Not performed in 4th C1.10 ALA2-3100-4R 10/11/2019 5th/1st/1R29 A

> 90%

no C1.10 ALA2-3100-5R Not performed in 4th C1.10 ALA2-3100-5R 10/10/2019 5th/1st/1R29 A

> 90%

no C1.20 ALA2-3300-6R 4/19/2006 3rd/3rd/1R20 A

> 90%

no C1.20 ALA2-3300-6R 10/14/2016 4th/3rd/1R27 A

> 90%

no C1.30 ALA2-3300-2R 4/12/2012 4th/2nd/1R24 A

> 90%

no C1.30 ALA2-3300-2R 4/2/2021 5th/2nd/1R30 A

> 90%

no A = Acceptable to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A1-5 Table 1-5. FNP2 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.40 APR1-3100-1R 4/17/2007 3rd/2nd/2R18 A

> 90%

no B2.40 APR1-3200-1R 4/22/2016 4th/3rd/2R24 A

> 90%

no C1.10 APR2-3100-4R Not performed in 4th C1.10 APR2-3100-4R 10/20/2023 5th/2nd/2FR29 A

> 90%

no C1.10 APR2-3100-5R Not performed in 4th C1.10 APR2-3100-5R 10/20/23 5th/2nd/2FR29 A

> 90%

no C1.20 APR2-3100-6R 11/5/2008 4th/1st/2FR19 A

> 90%

no C1.20 APR2-3100-6R 4/18/2019 5th/1st/2FR26 A

> 90%

no C1.30 APR2-3100-2R 11/1/2008 4th/1st/2FR19 A

> 90%

no C1.30 APR2-3100-2R 10/23/2017 4th/3rd/2FR25 A

> 90%

no A = Acceptable to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A1-6 Figure 1-1. FNP1&2 Steam Generator Layout and Item Nos.

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A1-7 References 1-1.

Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

1-2.

Structural Integrity Associates, Inc. Calculation No. FP-FNP-325, FatiguePro 4 Plant Farley Data Update - through 2022, Revision 0.

ATTACHMENT 2 PLANT-SPECIFIC APPLICABILITY VEGP UNITS 1 AND 2 to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-1 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 VEGP Units 1 and 2 is provided in Table 2-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to VEGP Units 1 and 2.

Table 2-1 Applicability of Reference [2-1] Representative Analyses to VEGP1&2 Item No. B2.40 (SG Primary Side Shell Welds)

Category Requirement from Reference [2-1]

Applicability to VEGP1&2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of a portion of the vessel) is not considered in this evaluation due to its rarity.

In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance.

VEGP1&2 have not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The VEGP1&2 SG channel heads are fabricated of SA-216 Gr. WCC material and the tubesheets are fabricated of SA-508 Cl. 2a material.

The limiting RTNDT value of the materials is 60°F which is consistent with the RTNDT of 60°F used in the EPRI report is conservative.

SA-508 Cl. 2a material conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Even though SA-216, Grade WCC is not specifically mentioned in ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110, this material is a low alloy pressure vessel steel with a specified minimum yield strength at room temperature of 50 ksi and has similar toughness properties to SA-533 and therefore conforms to the to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-2 Category Requirement from Reference [2-1]

Applicability to VEGP1&2 requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [2-1].

The VEGP1&2 weld configuration is shown in Figure 2-2 and conforms to that shown in Reference [2-1].

The SG vessel dimensions must be within 10%

of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [2-1].

Per Table 1 in the main section of this request for alternative, the VEGP1&2 SG vessel dimensions are as follows:

SG Lower Head OD = 137 SG Upper Shell OD = 175.92 The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [2-1] for Westinghouse plants.

The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [2-1] over a 60-year operating life.

As shown in Table 2-2, the VEGP1&2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-7 of Reference [2-1].

Items No. C1.10, C1.20 and C1.30 (SG Secondary Side Shell Welds)

Category Requirement from Reference [2-1]

Applicability to VEGP1&2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of a portion of the vessel) is not considered in this evaluation due to its rarity.

In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance.

VEGP1&2 have not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-3 Category Requirement from Reference [2-1]

Applicability to VEGP1&2 The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The VEGP1&2 SG vessel shells are fabricated of SA-533 Gr. A Cl. 2 material and tubesheets are fabricated of SA-508 Cl. 2a material.

The limiting RTNDT value of the materials is 60°F (so the RTNDT of 60°F used in the EPRI report is conservative).

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

Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [2-1].

The VEGP1&2 weld configurations are shown in Figures 2-3, 2-4 and 2-5 and conform to Figures 1-7 and 1-8 of Reference [2-1].

The SG vessel dimensions must be within 10%

of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [2-1].

Per Table 1 in the main section of this request for alternative, the VEGP1&2 SG vessel dimensions are as follows:

SG Lower Head OD = 137 SG Upper Shell OD = 175.92 The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [2-1] for Westinghouse plants.

The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [2-1] over a 60-year operating life.

As shown in Table 2-3, the VEGP1&2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-9 of Reference [2-1].

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-4 Table 2-2 VEGP1&2 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Welds (Comparison to Table 5-7 of Reference [2-1])

Transient(1)

Number of Cycles for 60 Years from Table 5-7 of Reference [2-1]

VEGP1 60-Year Projection (2)

VEGP2 60-Year Projection (3)

Heatup/Cooldown 300 67/64 73/71 Plant Loading/Unloading 5000 149/68 147/32 Reactor Trip 360 94 85 Notes:

1.

Table 5-7 of Reference [2-1] also includes allowable transient temperatures and pressures. From previous experience with Westinghouse plants, these values are with 2% of the values used in the EPRI report. This is acceptable based on the large margins in the evaluation as demonstrated by the low plant specific stress ratio compared to the maximum allowed stress ratio.

2.

From Table 11 of Reference [2-2].

3.

From Table 12 of Reference [2-2].

Table 2-3 VEGP1&2 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Welds (Comparison to Table 5-9 of Reference [2-1])

Transient(1)

Number of Cycles for 60 Years from Table 5-9 of Reference [2-1]

VEGP1 60-Year Projection (2)

VEGP2 60-Year Projection (3)

Heatup/Cooldown 300 67/64 73/71 Plant Loading/Unloading 5000 149/68 148/32 Reactor Trip 360 94 85 Notes:

1.

Table 5-9 of Reference [2-1] also includes allowable transient temperatures and pressures. From previous experience with Westinghouse plants, these values are with 2% of the values used in the EPRI report. This is acceptable based on the large margins in the evaluation as demonstrated by the low plant specific stress ratio compared to the maximum allowed stress ratio.

2.

From Table 11 of Reference [2-2].

3.

From Table 12 of Reference [2-2].

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-5 Table 2-4. VEGP1 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.40 11201-B6-001-W08 3/20/2011 3rd/2nd/1R16 A

80%

RR-03 B2.40 11201-B6-001-W08 10/3/2000 2nd/2nd/1R9 A

80%

RR-6 C1.10 11201-B6-002-W03 3/17/2020 4th/1st/1R22 A

<90 Submittal will follow 4th interval C1.10 11201-B6-002-W03 10/3/2009 3rd/2nd/1R15 A

>90 N/A C1.10 11201-B6-002-W04 3/26/2014 3rd/3rd/1R18 A

>90 N/A C1.10 11201-B6-002-W04 3/25/2005 2nd/3rd/1R12 A

>90 N/A C1.20 11201-B6-001-W01 9/23/2018 4th/1st/1R21 A

>90 N/A C1.20 11201-B6-001-W01 3/27/2008 3rd/1st/1R14 A

>90 N/A C1.30 11201-B6-001-W007 3/7/2023 4th/2nd/1R24 A

>90 N/A C1.30 11201-B6-004-W007 3/19/2011 3rd/2nd/1R16 A

>90 N/A A = Acceptable to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-6 Table 2-5. VEGP2 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.40 21201-B6-004-W08 9/18/2014 3rd/2nd/2R17A A

80%

RR-3 B2.40 21201-B6-004-W08 5/1/2004 2nd/2nd/1R10 A

80%

RR-6 C1.10 21201-B6-002-W04 9/26/2014 3rd/ 3rd/ 2R17 A

>90 N/A C1.10 21201-B6-002-W04 9/28/2005 2nd/3rd/2R11 A

>90 N/A C1.10 21201-B6-002-W03 9/19/2023 4rd/2nd/2R23 A

>90 N/A C1.10 21201-B6-002-W03 3/16/2010 3rd/1st//2R14 A

>90 N/A C1.20 21201-B6-001-W01 3/18/2019 4th/1st/2R20 A

>90 N/A C1.20 21201-B6-001-W01 9/27/2008 3rd/1st/2R13 A

>90 N/A C1.30 21201-B6-004-W007 9/16/2023 4th/2nd/2R23 A

>90 N/A C1.30 21201-B6-004-W007 9/30/2011 3rd/2nd/2R15 A

>90 N/A A = Acceptable to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-7 Figure 2-1. VEGP1&2 Steam Generator Layout (Reference [2-3])

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A2-8 References 2-1.

Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

2-2.

Structural Integrity Associates, Inc. Calculation No. FP-FNP-332, FatiguePro 4 Baseline Analysis of Vogtle Plant Cycles and Fatigue Usage - Startup through 12/31/2022, Revision 0.

2-3.

Drawing No. ISI-1 1201-B6-001, Vogtle Electric Generating Plant Unit - 1 Inservice Inspection Eqpt. Dwg. Steam Generator - System 1201, Revision 5.

ATTACHMENT 3 RESULTS OF INDUSTRY SURVEY to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A3-1 Overall Industry Inspection Summary for Code Items B2.31, B2.32, B2.40, B3.130, C1.10, C1.20, and C1.30 The results of an industry survey of past inspections of SG nozzle-to-shell welds, inside radius sections and shell welds are summarized in Reference [3-1]. Table 3-1 provides a summary of the combined survey results for Item Nos. B2.31, B2.32 (see Table Note 3), B.240, B3.130, C1.10, C1.20, and C1.30. The results of the industry survey identified numerous steam generator (SG) examinations being performed with no service-induced flaws being detected. Performing these examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international boiling water reactor (BWR) and pressurized water reactor (PWR) units responded to the survey and provided information representing all PWR plant designs currently in operation in the United States. 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 1374 examinations for the components of the affected Item Nos. were conducted, with 1148 of these specifically for PWR components. The majority of PWR examinations were performed on SG welds.

A relatively small number of flaws were identified during these examinations which required flaw evaluation. None of these flaws were found to be service-induced. For Item No. B2.40, examinations at two units at a single plant site identified multiple flaws exceeding the acceptance criteria of ASME Code Section XI; however, these were determined to be subsurface-embedded fabrication flaws and non-service-induced (see Table Note 1). For Item No. C1.20, two PWR units reported flaws exceeding the acceptance criteria of ASME Code,Section XI. In the first unit, a single flaw was identified, and was evaluated as an inner diameter surface imperfection. Reference [3-3] indicates that this was a spot indication with no measurable through-wall depth. This indication is therefore not considered to be service-induced but rather fabrication-related.

A flaw evaluation per IWC-3600 was performed for this flaw and it was found to be acceptable for continued operation. In the second unit, multiple flaws were identified (see Table Note 2). As discussed in References [3-4] and [3-5], these flaws were most likely subsurface weld defects typical of thick vessel welds and not service-induced. A flaw evaluation for IWC-3600 was performed for these flaws and they were found to be acceptable for continued operation.

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A3-2 Table 3-1. Summary of Survey Results for SG Nozzle-to-Shell, Inside Radius Section, and Shell Weld Components Item No.

No. of Examinations No. of Reportable Indications BWR PWR Total BWR PWR Total B2.31 0

30 30 0

0 0

B2.32 (Note 3) 0 13 13 0

0 0

B2.40 0

183 183 0

Note 1 Note 1 B3.130 0

135 135 0

0 0

C1.10 140 305 445 0

0 0

C1.20 54 319 373 0

Note 2 Note 2 C1.30 32 163 195 0

0 0

Totals 226 1148 1374 0

Notes 1 and 2 Notes 1 and 2 Notes:

1. Two PWR W-2 Loop units at a single plant reported multiple subsurface embedded fabrication flaws.

2. A single PWR W-2 Loop unit reported multiple flaws (References [3-4 and 3-5]).

3. Item No. B2.32 was evaluated in the Reference [3-1] technical basis and included in the industry survey, but is not contained in the scope of this alternative request.

to NL-24-0201 Proposed Alternative GEN-ISI-ALT-2024-02 in Accordance with 10 CFR 50.55a(z)(1)

A3-3 References 3-1.

Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

3-2.

Not used.

3-3.

Letter from F. A. Kearney (Exelon) to NRC, Byron Station Unit 2 90-Day Inservice Inspection Report for Interval 3, Period 3, (B2R17), dated July 29, 2013, ADAMS Accession Number ML13217A093.

3-4.

Letter from J. M. Sorensen (NMC) to NRC, Unit 1 Inservice Inspection Summary Report, Interval 3, Period 3 Refueling Outage Dates 1-19-2001 to 2-25-2001 Cycle 20 /

05-26-99 to 02-25-2001, dated May 29, 2001, ADAMS Accession Number ML011550346.

3-5.

Letter from J. P. Solymossy (NMC) to NRC, Response to Opportunity For Comment On Task Interface Agreement (TIA) 2003-01, Application of ASME Code Section XI, IWB-2430 Requirements Associated With Scope of Volumetric Weld Expansion at the Prairie Island Nuclear Generating Plant, dated April 4, 2003, ADAMS Accession Number ML031040553.