0CAN062402, Proposed Alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles

Proposed Alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles

ML24158A389
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/06/2024
From:	Couture P Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
0CAN062402
Download: ML24158A389 (1)

Text

Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213 0CAN062402 10 CFR 50.55a June 6, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Proposed Alternatives ANO1-ISI-24-01 and ANO2-ISI-24-01 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles Arkansas Nuclear One - Units 1 and 2 NRC Docket Nos. 50-313 and 50-368 Renewed Facility Operating License Nos. DPR-51 and NPF-6 In accordance with 10 CFR 50.55a(z)(1), Entergy Operations, Inc. (Entergy) hereby requests Nuclear Regulatory Commission (NRC) approval of proposed alternatives for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2, respectively).

The proposed alternatives are to defer the In-Service Inspection (ISI) examinations for select examination categories and item numbers for the steam generators at ANO-1 and ANO-2 from the current American Society of Mechanical Engineers (ASME) Code,Section XI, Division 1 10-year requirements to the end of currently licensed operating life, which is scheduled to end on May 20, 2034, and July 17, 2038, respectively. This equates to an extension of 16 years, 11 months, 20 days from the end of the fourth ISI interval (May 30, 2017) for ANO-1 and an extension of 17 years, 3 months, 23 days from the end of the fourth ISI interval (March 25, 2021) for ANO-2 at which time all ASME Code,Section XI, Division 1 requirements were satisfied. Entergy requests authorization to use the proposed alternatives pursuant to 10 CFR 50.55a(z)(1) on the basis that the alternative provides an acceptable level of quality and safety.

The Enclosure to this letter provides the Alternative Requests ANO1-ISI-24-01 and ANO2-ISI-24-01 with Enclosure, Attachment 1 providing the Plant-Specific Applicability ANO-1, Enclosure, Attachment 2 providing the Plant-Specific Applicability ANO-2, and Enclosure, providing the Results of Industry Survey.

Entergy requests approval of the proposed Alternative Requests by May 30, 2025. This date will support the conclusion of the second period, fifth ISI interval for ANO-1.

Phil Couture Senior Manager Fleet Regulatory Assurance - Licensing 601-368-5102

0CAN062402 Page 2 of 2 This letter contains no new regulatory commitments.

If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, Arkansas Nuclear One, at 479-858-7826.

Respectfully, Phil Couture PC/rwc

Enclosure:

Alternative Requests ANO1-ISI-24-01 and ANO2-ISI-24-01 Attachments to

Enclosure:

1. Plant-Specific Applicability ANO-1

2. Plant-Specific Applicability ANO-2

3. Results of Industry Survey cc:

NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Digitally signed by Philip Couture DN: cn=Philip Couture, c=US, o=Entergy, ou=Regulatory Assurance, email=pcoutur@entergy.com Date: 2024.06.06 13:11:38 -05'00' Philip Couture

ENCLOSURE 0CAN062402 ALTERNATIVE REQUESTS ANO1-ISI-24-01 AND ANO2-ISI-24-01

0CAN062402 Enclosure Page 1 of 24 ALTERNATIVE REQUESTS ANO1-ISI-24-01 AND ANO2-ISI-24-01 ASME CODE COMPONENTS AFFECTED:

Code Class:

Class 1 and Class 2

==

Description:==

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

Examination Category: Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 1, Category B-D, full penetration welded nozzles in vessels Class 2, Category C-A, pressure-retaining welds in pressure vessels Class 2, Category C-B (Pressure Retaining Nozzle Welds in Pressure Vessels,Section XI, Division 1)

Item Numbers:

B2.31 - Steam Generator (Primary side), head welds, circumferential B2.40 - Steam Generators (primary side), tube sheet-to-head weld B3.130 - Steam Generators (primary side), nozzle-to-vessel welds C1.20 - Head Circumferential Welds C1.30 - Tube sheet-to-shell weld C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections

0CAN062402 Enclosure Page 2 of 24 ANO Unit 1 (ANO-1)

ASME Category ASME Item No.

Component ID Component Description B-B B2.40 03-103 Tube sheet-to-head weld B-B B2.40 03-109 Tube sheet-to-head weld B-B B2.40 04-103 Tube sheet-to-head weld B-B B2.40 04-109 Tube sheet-to-head weld B-B B2.31 03-102 Head weld circumferential B-B B2.31 03-110 Head weld circumferential B-B B2.31 04-102 Head weld circumferential B-B B2.31 04-110 Head weld circumferential C-A C1.30 03-104 Tube sheet-to-shell welds C-A C1.30 03-108 Tube sheet-to-shell welds C-A C1.30 04-104 Tube sheet-to-shell welds C-A C1.30 04-108 Tube sheet-to-shell welds C-B C2.21 03-117 Nozzle-to-shell welds Main Steam (MS)

C-B C2.21 03-118 Nozzle-to-shell welds (MS)

C-B C2.21 04-117 Nozzle-to-shell welds (MS)

C-B C2.21 04-118 Nozzle-to-shell welds (MS)

C-B C2.22 03-115IR Nozzle inside radius sections (MS)

C-B C2.22 03-116IR Nozzle inside radius sections (MS)

C-B C2.22 04-115IR Nozzle inside radius sections (MS)

C-B C2.22 04-116IR Nozzle inside radius sections (MS)

0CAN062402 Enclosure Page 3 of 24 ANO Unit 2 (ANO-2)

ASME Category ASME Item No.

Component ID Component Description B-B B2.40 03-004 Tube sheet-to-head welds B-B B2.40 04-004 Tube sheet-to-head welds B-D B3.130 03-005 Nozzle-to-vessel welds B-D B3.130 04-005 Nozzle-to-vessel welds B-D B3.130 03-006 Nozzle-to-vessel welds B-D B3.130 04-006 Nozzle-to-vessel welds B-D B3.130 03-007 Nozzle-to-vessel welds B-D B3.130 04-007 Nozzle-to-vessel welds C-A C1.20 03-001 Head circumferential welds C-A C1.20 04-001 Head circumferential welds C-A C1.30 03-003 Tube sheet-to-shell welds C-A C1.30 04-003 Tube sheet-to-shell welds C-B C2.21 03-002 Nozzle-to-shell welds Feedwater (FW)

C-B C2.21 04-002 Nozzle-to-shell welds (FW)

C-B C2.22 03-002IR Nozzle inside radius sections (FW)

C-B C2.22 04-002IR Nozzle inside radius sections (FW)

APPLICABLE CODE EDITION AND ADDENDA:

ANO-1 The fifth 10-year in-service inspection (ISI) interval Code of record for ANO-1 is the 2007 Edition of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,Section XI through the 2008 Addenda, "Rules for Inservice Inspection of Nuclear Power Plant Components."

ANO-2 The fifth 10-year ISI interval Code of record for ANO-2 is the 2007 Edition of the ASME B&PV Code,Section XI through the 2008 Addenda, "Rules for Inservice Inspection of Nuclear Power Plant Components.

0CAN062402 Enclosure Page 4 of 24 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 Categories C-A and C-B require examination of the following Item Nos.:

Item No. B2.31 Volumetric of all nozzles during the first Section XI inspection interval and one weld per head during successive intervals. The examination areas are shown in Figures IWB-2500-3.

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. B3.130 Volumetric examination of all nozzles during each Section XI inspection interval. The examination areas for are shown in Figures IWB-2500-7(a), (b),

(c) and (d).

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

Item No. C1.30 Volumetric examination of essentially 100% of the weld length of the tube sheet-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.

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

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

0CAN062402 Enclosure Page 5 of 24 REASON FOR REQUEST:

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

[9.2] of the bases for the ASME Code,Section XI examination requirements specified for the above listed ASME Code,Section XI, Division 1 examination categories for SG welds and components. 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] and [9.2] reports 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 two EPRI reports supplemented by plant-specific evaluations contained herein, Entergy is requesting an ISI examination deferral for the subject welds. The Reference [9.1] and [9.2] reports were developed consistent with the recommendations provided in EPRIs 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].

PROPOSED ALTERNATIVE AND BASIS FOR USE:

ANO-1 For ANO-1, Entergy 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 B2.31 Steam generators (primary side), head welds, circumferential B-B B2.40 Steam generators (primary side), tube sheet-to-head weld C-A C1.30 Tube sheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C-B C2.22 Nozzle inside radius sections In 2005 (third period of the third inspection, interval) the ANO-1 SGs were replaced. The new SG welds and components received the required pre-service inspection (PSI) examinations followed by ISI examinations through the second period of the current fifth inspection interval.

The proposed alternative is to defer the ISI examinations for these Item Nos. for the SGs at ANO-1 from the current ASME Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is currently scheduled to end on May 20, 2034. This equates to an extension of 16 years, 11 months, 20 days from the end of the fourth ISI interval (May 30, 2017) at which time all ASME Code,Section XI, Division 1 requirements were satisfied.

0CAN062402 Enclosure Page 6 of 24 ANO-2 For ANO-2, Entergy 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), tube sheet-to-head weld B-D B3.130 Steam generators (primary side), nozzle-to-vessel welds C-A C1.20 Head circumferential weld C-A C1.30 Tube sheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C-B C2.22 Nozzle inside radius sections In 2000 (first period of the third inspection interval), the ANO-2 SGs were replaced. The new SG welds and components received the required PSI examinations 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 SGs at ANO-2 from the current ASME Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is currently scheduled to end on July 17, 2038. This equates to an extension of 17 years, 3 months, 23 days from the end of the fourth ISI interval (March 25, 2021) at which time all ASME Code,Section XI, Division 1 requirements were satisfied.

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 ANO-1 and 2 is shown in Attachments 1 and 2.

Applicability of the Degradation Mechanism Evaluation in References [9.1] and [9.2] to ANO-1 and 2 An evaluation of degradation mechanisms that could potentially impact the reliability of the SG welds and components was performed in References [9.1] and [9.2]. 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

0CAN062402 Enclosure Page 7 of 24 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 Vogtle 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 the plants considered in this Request for Alternative are like those in the References [9.1] and [9.2] and therefore, the conclusions of these reports apply to the ANO-1 and 2. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in References [9.1] and [9.2].

As part of the technical basis in References [9.1] and [9.2], a comprehensive industry survey involving 74 Pressurized Water Reactor (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 References [9.1] and [9.2] to ANO-1 and 2 Finite element analyses (FEA) were performed in References [9.1] and [9.2] to determine the stresses in the SG welds and components covered in this request. The finite element models used in References [9.1] and [9.2] are consistent with the configurations of ANO-1 and 2 and therefore no new FEA model is required for the stress analysis of these plants. The analysis in References [9.1] and [9.2] was performed using representative 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 ANO-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] and [9.2] stress analyses are compared to those of ANO-1 and 2 in Tables 1 and 2:

0CAN062402 Enclosure Page 8 of 24 Table 1 SG Vessel Dimensions Plant Primary Lower Head Inside Diameter (ID) (in)

Primary Lower Head Thickness (in)

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

Secondary Upper Shell Thickness (in)

Secondary Upper Shell Ri/t EPRI Report (Table 4-2 of

[9.2])

155.33 6.94 11.2 230.87 4.91 23.5 ANO-1 118.6(2) 8.2(2) 7.23 139.95(1) 5.4(1) 13.0 ANO-2 157.42(5) 6.94(3) 11.34 230.87(4) 4.79(4) 24.1 Notes:

1. Reference [9.21]

2. Reference [9.22]

3. Reference [9.23]

4. Reference [9.24]

5. Reference [9.33]

0CAN062402 Enclosure Page 9 of 24 Table 2 SG Nozzle Dimensions Plant FW Nozzle ID (in)

FW Nozzle Thickness (in)

FW Nozzle Ri/t MS Nozzle ID (in)

MS Nozzle Thickness (in)

MS Nozzle Ri/t EPRI Report (Figures 4-9 and 4-10 of [9.1])

16.5 6

1.38 22.25 4.53 2.46 ANO-1 N/A(1)

N/A(1)

N/A(1) 21.75(3) 5.31(4) 2.05 ANO-2 18.0(2) 7.25(2) 1.24 30.8 8.11 1.90 Notes:

1. Nozzle has no C2.21 or C2.22 components.

2. Reference [9.30]

3. Reference [9.34]

4. Reference [9.35]. This thickness value is for the shell which (per Reference [9.36]) is like the MS nozzle thickness.

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 SE for Vogtle [9.16], the dominant stress is the pressure stress. Therefore, the variation in the Ri/t ratio determined in Tables 1 and 2 can be used to scale up the stresses of the Reference [9.1] and [9.2] reports to obtain the plant-specific stresses for each unit and component. From Tables 1 and 2, the stress ratio (Ri/t) of ANO-1 and ANO-2 relative to that used in the EPRI report) are as follows.

Primary lower head: ANO-1 (7.23/11.2) = 0.65 and ANO-2 (11.34/11.2) = 1.01 (applicable to primary side welds but conservatively assumed applicable to the rest of the SG welds)

Secondary upper shell: ANO-1 (13/23.5) = 0.55 and ANO-2 (24.1/23.5) = 1.03 (applicable to FW and MS nozzle-to-shell welds)

FW nozzle: ANO-2 (1.24/1.38) = 0.90 (applicable to FW inside radius sections) (Note: N/A for ANO-1 per Table 2)

MS nozzle: ANO-1 (2.05/2.46) = 0.83 and ANO-2 (1.90/2.46) = 0.92 (applicable to MS inside radius sections)

In the selection of the transients in Section 5 of References [9.1] and [9.2] and the subsequent stress analyses in Section 7, test conditions beyond a system leakage test were not considered since pressure tests at ANO-1 and 2 are performed at normal operating conditions. No hydrostatic testing had been performed at ANO-1 and 2 since the units went into operation.

0CAN062402 Enclosure Page 10 of 24 In Reference [9.2], 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 [9.2] and this was found acceptable by the NRC in Section 5.3 of Reference [9.18].

Applicability of the Flaw Tolerance Evaluation in References [9.1] and [9.2] to ANO-1 and 2 Flaw tolerance evaluations were performed in References [9.1] and [9.2] consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a PSI followed by subsequent ISI, the NRCs safety goal of 1.0E-06 failures per year is met.

The PFM analysis in Reference [9.1] was performed using the Probabilistic OptiMization of InSpEction (PROMISE) Version 1.0 software, developed by Structural Integrity Associates. As part of the NRCs review of Southern Nuclears alternative request, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan Reference [9.37].

The PFM analysis in Reference [9.2] 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.

In Section 8.2.2.2 of Reference [9.1] and Section 8.3.2.2 of Reference [9.2], a nozzle flaw density of 0.001 flaws per nozzle was assumed for the nozzle inside radius sections. In Section 3.8.5 of the SE for Vogtle in Reference [9.14], the NRC indicated that a nozzle flaw density of 0.1 flaws per nozzle should have been used. Sensitivity studies performed in Section 8.2.4.3.4 in Reference [9.2] indicated that by changing the number of flaws in the nozzle inside radius sections from 0.001 to 0.1, the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1.0E-06 per year. A comparison of the PSI/ISI scenarios used in the sensitivity studies performed in References [9.1] and [9.2] to those at ANO-1 and 2 is provided below. Note that the assumption below of a 30-year ISI deferral is conservative compared to the end of currently licensed operating life for each plant.

ANO-1 For the ANO-1 replacement SGs installed in 2005 (third period of the third inspection interval),

PSI examinations have been performed followed by ISI examinations in the one completed 10-year interval (fourth interval) following SG replacement. The PSI/ISI scenario considered is therefore PSI plus one 10-year ISI examinations to be followed by two 30-year ISI deferrals (PSI+10+40+70).

0CAN062402 Enclosure Page 11 of 24 ANO-2 For the ANO-2 replacement SGs installed in 2000 (first period of the third inspection interval),

PSI examinations have been performed followed by ISI examinations in the two completed 10-year intervals (third and fourth intervals) following SG replacement. The PSI/ISI scenario considered is therefore PSI plus two 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+50).

Limiting PSI/ISI Scenario From Reference [9.1], the limiting component for Item Nos. C2.21 and C2.22 is the FW nozzle.

However, there are no Item No. C2.21 and C2.22 components for the ANO-1 Babcock & Wilcox (B&W) SG FW nozzle configuration; therefore, the MS nozzles will be considered at ANO-1.

Five separate evaluations are performed with the following limiting PSI/ISI scenarios:

1. ANO-1 MS nozzle inside radius sections (PSI+10+40+70)

2. ANO-1 MS nozzle-to-shell welds PSI+10+40+70)

3. ANO-2 FW nozzle inside radius sections (PSI+10+20+50)

4. ANO-2 FW nozzle to-shell welds (PSI+10+20+50)

5. The remainder of the SG welds (applicable to both ANO-1 and ANO-2) (PSI+10+40+70)

The above limiting PSI/ISI scenarios for ANO-1 and 2 were not specifically considered in the Reference [9.1] and [9.2] PFM evaluations in combination with key variables, as evaluated by the NRC in Section 4.0 (page 6) of the Reference [9.16] SE. Therefore, the following additional plant-specific evaluations were performed with the limiting PSI/ISI scenarios shown above.

ANO-1 MS Nozzle Inside Radius Section From Reference [9.1], the critical Case ID for the MS nozzle inside radius section is SGB-P1N.

An evaluation like that shown in Table 8-28 of Reference [9.1] was performed for this location assuming a nozzle flaw density of 0.1, a fracture toughness of 200 kilo pound per square inch square root inch (ksiin) and a standard deviation 5 ksiin as described by the NRC in Reference [9.13]. A relatively high stress multiplier of 2.35 was applied to get close to the acceptance criteria. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 3 and show that after 80 years of plant operation the probabilities of rupture and leakage are well below the acceptance criterion of 1.0E-06.

0CAN062402 Enclosure Page 12 of 24 Table 3 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for B&W MS Nozzle Inside Radius Section (Case ID SGB-P1N from Reference [9.1])

Time (yr)

Probability per Year for Combined Case KIC = 200 ksiin.

SD = 5 ksiin.

Stress Multiplier = 2.35 Nozzle Flaw Density = 0.1 PSI+10+40+70 Rupture Leak 10 1.00E-09 1.00E-09 20 5.00E-10 5.00E-10 30 2.83E-08 2.00E-09 40 3.26E-07 1.61E-07 50 2.61E-07 1.30E-07 60 2.20E-07 1.12E-07 70 1.92E-07 1.01E-07 80 1.68E-07 8.88E-08 ANO-1 MS Nozzle-to-Shell Weld For the MS nozzle-to-shell weld, Table 8-15 of Reference [9.1] indicates that the critical Case ID is SGB-P3A. For the evaluation, a flaw density of 1.0 flaw per weld was assumed, consistent with the evaluations in Reference [9.1]. A fracture toughness of 200 ksiin and standard deviation of 5 ksiin were also used. A relatively high stress multiplier of 1.95 was applied. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 4 and show that after 80 years of plant operation the probabilities of rupture and leakage are well below the acceptance criterion of 1.0E-06. The results indicate that a much higher stress multiplier than 1.95 could have been used and the acceptance criteria would still be met.

0CAN062402 Enclosure Page 13 of 24 Table 4 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the B&W MS Nozzle-to-Shell Weld Case ID SGB-P3A from Reference [9.1])

Time (year)

Probability per Year for Combined Case KIC = 200 ksiin.

SD = 5 ksiin.

Stress Multiplier = 1.95 Nozzle Flaw Density = 1 PSI+10+40+70 Rupture Leak 10 1.00E-08 1.00E-08 20 5.00E-09 5.00E-09 30 2.00E-08 3.33E-09 40 7.20E-07 2.50E-09 50 5.80E-07 2.00E-09 60 5.00E-07 1.67E-09 70 4.80E-07 1.43E-09 80 4.20E-07 1.25E-09 ANO-2 FW Nozzle Inside Radius Section From Reference [9.1], the critical location for the inside radius section is FW nozzle Case ID FEW-P1N. An evaluation like that shown in Table 8-28 of Reference [9.1] was performed for this location assuming a nozzle flaw density of 0.1, a fracture toughness of 200 ksiin and a standard deviation 5 ksiin as recommended by the NRC in Reference [9.14]. A stress multiplier of 1.75 was applied. This stress multiplier was chosen to result in probability of rupture or probability of leakage close to the acceptance criteria after 80 years. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 5 and show that after 80 years of plant operation the probabilities of rupture and leakage are below the acceptance criterion of 1.0E-06.

0CAN062402 Enclosure Page 14 of 24 Table 5 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for Westinghouse (bounds CE design) FW Nozzle Inside Radius Section (Case ID FEW-P1N from Reference [9.1])

Time (yr)

Probability per Year for Combined Case KIC = 200 ksiin.

SD = 5 ksiin.

Stress Multiplier = 1.75 Nozzle Flaw Density = 0.1 PSI+10+20+50 Rupture Leak 10 3.97E-07 1.58E-07 20 2.41E-07 9.80E-08 30 1.60E-07 6.57E-08 40 1.22E-07 4.98E-08 50 1.06E-07 4.26E-08 60 8.85E-08 3.55E-08 70 7.60E-08 3.04E-08 80 6.65E-08 2.66E-08 ANO-2 FW Nozzle-to-Shell Weld For the FW nozzle-to-shell weld, Table 8-16 of Reference [9.1] indicates that the critical Case ID is FEW-P3A. For the evaluation, a nozzle flaw density of 1 flaw per nozzle was assumed. A fracture roughness of 200 ksiin and standard deviation 5 ksiin were also used. A stress multiplier of 1.45 was applied such that probability of rupture or probability of leakage are close to the acceptance criteria after 80 years. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 6 and show that after 80 years of plant operation the probabilities of rupture and leakage are below the acceptance criterion of 1.0E-06.

0CAN062402 Enclosure Page 15 of 24 Table 6 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for Westinghouse (bounds CE design) FW Nozzle-to-Shell Weld (Case ID FEW-P3A from Reference [9.1])

Time (year)

Probability per Year for Combined Case KIC = 200 ksiin.

SD = 5 ksiin.

Stress Multiplier = 1.45 Nozzle Flaw Density = 1 PSI+10+20+50 Rupture Leak 10 1.00E-08 1.00E-08 20 5.00E-09 1.00E-08 30 3.33E-09 6.67E-09 40 2.50E-09 3.75E-08 50 2.00E-09 7.48E-07 60 1.67E-09 6.30E-07 70 1.43E-09 5.63E-07 80 1.25E-09 5.61E-07 Remainder of ANO-1 and 2 SG Welds For the remaining SG welds, Table 8-32 of Reference [9.2] indicates that the critical Case ID is SGPTH-P4A. This case was evaluated for the limiting inspection scenario of PSI+10+40+70, 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 1.8 was applied. The results of the evaluation, using PROMISE Version 2.0, are summarized in Table 7 and show that after 80 years of plant operation, the probabilities of rupture and leakage are well below the acceptance criterion of 1.0E-06. The results indicate that a much higher stress multiplier than 2.1 could have been used and the acceptance criteria would still be met.

0CAN062402 Enclosure Page 16 of 24 Table 7 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the Remaining SG Welds (CE or B&W)

(Case ID SGPTH-P4A from Reference [9.2])

Time (year)

Probability per Year for Combined Case KIC = 200 ksiin.

SD = 5 ksiin.

Stress Multiplier = 1.8 Nozzle Flaw Density = 1 PSI+10+40+70 Rupture Leak 10 1.00E-08 1.00E-08 20 5.00E-09 5.00E-09 30 2.33E-08 3.33E-09 40 2.33E-07 2.50E-09 50 1.86E-07 2.00E-09 60 1.70E-07 1.67E-09 70 1.71E-07 1.43E-09 80 1.50E-07 1.25E-09 The plant-specific PFM evaluation presented above for ANO-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 1.0xE-06 failures per year. The stress multipliers applied Tables 3 through 7 are greater than the plant specific stress ratios determined previously from the geometrical data in Tables 1 and 2 and therefore the stresses and fracture mechanics evaluations in the References [9.1] and [9.2] EPRI reports are conservative in application to ANO-1 and 2. It should also be noted that the evaluation incorporates conservative assumptions about the PSI/ISI scenarios. Furthermore, the evaluation was performed for 30 years, which is longer than the deferral being sought by Entergy in this Request for Alternative.

An evaluation was performed to show acceptability of the low KIC values at the beginning and ending of the heat up/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 Branch Technical Position (BTP) 5-3 [9.26]. The RTNDT value of 60oF is consistent with the limiting value assumed in Attachment 2 for the SG materials at ANO-2. 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

0CAN062402 Enclosure Page 17 of 24 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 References [9.1] and [9.2] 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. However, for Class 2 components, the use of Appendix VIII procedures is plant specific. In the case of ANO-1 and 2, ASME Code,Section V procedures may be used for at least some Class 2 components (e.g., the ANO-1 MS nozzle inside radius section examinations and the ANO-2 FW nozzle inside radius section examinations). 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 8-31 of Reference [9.1] and Table 8-3 of Reference [9.2] provide verification of the above PFM results for ANO-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 8.2.4.1.1 of Reference [9.1] and Section 8.3.4.1 of Reference [9.2], PSI examination 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.

ANO-1 Inspection history for ANO-1 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 1. As shown in the attachment, one weld/component has limited examination coverage of 73.75%. Examination coverage greater than 50% is acceptable per Section 3.8.7 of the Vogtle SE [9.16]. As shown in, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

ANO-2 Inspection history for ANO-2 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 2. As shown in the attachment, one weld/component has limited exam coverage of 69.60%. Examination coverage greater than 50% is acceptable per Section 3.8.7 of the Vogtle SE [9.16]. As shown in

0CAN062402 Enclosure Page 18 of 24

, 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. The results of the survey indicate that these components are very flaw tolerant.

Conclusion It is concluded that the SG pressure-retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis reports

[9.1] and [9.2], 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 10-6 failures per reactor year is met with considerable margins.

Plant-specific applicability of the technical basis to ANO-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 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 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, ANO-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, Entergy requests the NRC grant this proposed alternative in accordance with 10 CFR 50.55a(z)(1).

DURATION OF PROPOSED ALTERNATIVE:

ANO-1 The proposed alternative is requested for the remainder of the current fifth inspection interval and through the end of currently licensed operating life, which is currently scheduled to end on May 20, 2034.

0CAN062402 Enclosure Page 19 of 24 ANO-2 The proposed alternative is requested for the remainder of the current fifth inspection interval and through the end of currently licensed operating life, which is currently scheduled to end on July 17, 2038.

PRECEDENTS:

The following previous submittal has been made by Southern Nuclear 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 (Southern Nuclear) to the NRC, 'Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0," ADAMS Accession No. ML20253A311. dated September 9, 2020, [9.15].

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

Letter from Michael T. Markley (NRC) 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), ADAMS Accession No. ML20352A155, dated January 11, 2021,

[9.16].

The following previous submittal has been 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, ADAMS Accession No. ML20198M682, dated July 15, 2020 [9.17].

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

Letter from James G. Danna (NRC) 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), ADAMS Accession No. ML21167A355, dated July 16, 2021 [9.18].

0CAN062402 Enclosure Page 20 of 24 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 (TAC No. MA 5446), ADAMS Accession No. ML003730922, dated July 24, 2000.

Letter from R. L. Emch (NRC) to J. B. Beasley, Jr. (Sothern Nuclear Operating Company), 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 (TAC No.

MB0603 and MB0604), ADAMS Accession No. ML011640178, dated June 20, 2001.

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

Letter from M, Khanna (NRC) to D. A. Heacock (Dominion Nuclear Connecticut Inc.),

Millstone Power Plant Unit No. 2 - Issuance of Relief Requests RR-89-69 Through RR-89-78 Regarding Third 10-Year Interval Inservice Inspection plan (TAC Nos. ME5998 Through ME6006), ADAMS Accession No. ML120541062, dated March 12, 2012.

Letter from R. J. Pascarelli (NRC) to E. D. Halpin (Pacific Gas & Electric), 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 (CAC Nos. MF6646 and MF6647), ADAMS Accession No. ML15337A021, dated December 8, 2015.

In addition, there are precedents related to similar topical 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 Boiling Water Reactor Vessel and Internals Program (BWRVIP)-108 [9.6] and BWRVIP-241 [9.8], the NRC approved the reduction of Boiling Water Reactor (BWR) vessel FW 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 19 of Regulatory Guide 1.147 [9.11].

0CAN062402 Enclosure Page 21 of 24 ACRONYMS:

ASME American Society of Mechanical Engineers B&W Babcock and Wilcox BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Program CE Combustion Engineering CFR Code of Federal Regulations DFM Deterministic fracture mechanics EAF Environmentally assisted fatigue EPRI Electric Power Research Institute FAC Flow accelerated corrosion FEA Finite element analysis FW Feedwater ISI Inservice Inspection MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system O.D.

Outside diameter PDI Probability of detection PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor SCC Stress corrosion cracking SG Steam Generator WEC Westinghouse Electric Company

REFERENCES:

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

2019. 3002014590.

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

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.

0CAN062402 Enclosure Page 22 of 24 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, ADAMS Accession No. ML111600303, July 26, 2011.

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),

ADAMS Accession No. ML073600374, dated December 19, 2007.

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

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, ADAMS Accession No. ML19241A545, February 27, 2019.

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

9.16 Letter from Michael T. Markley (NRC) 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), ADAMS Accession No. ML20352A155, dated January 11, 2021.

0CAN062402 Enclosure Page 23 of 24 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, ADAMS Accession No. ML20198M682, dated July 15, 2020.

9.18 Letter from James G. Danna (NRC) 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), ADAMS Accession No. ML21167A355, dated July 16, 2021.

9.19 Electronic mail Letter from R. Guzman (NRC) to S. Sinha (Dominion Energy Nuclear Connecticut, Inc.), Millstone Unit 2 - Request for Additional Information - Alternative Request RR-05-06 Inspection Interval Extension for SG Pressure Retaining Welds and Full-Penetration Welded Nozzles (EPID: L-2020-LLR-0097), ADAMS Accession No. ML21034A576, dated February 3, 2021.

9.20 Letter from G. T. Bischof (Dominion Energy Nuclear Connecticut, Inc.) to the NRC, Dominion Energy Nuclear Connecticut, Inc., Millstone Power Station Unit 2 - Response to Request for Additional Information for Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure Retaining Welds and Full-Penetration Welded Nozzles, ADAMS Accession No. ML21081A136, dated March 19, 2021.

9.21 UT Vessel Examination Report No. 1-ISI-UT-11-007 (file 1R23 03-104 UT.pdf),

10/27/11.

9.22 UT Vessel Examination Report No. 1-ISI-UT-11-014 (file 1R23 03-103 UT.pdf),

10/27/11.

9.23 Drawing No. M-2001-C6-293, Arkansas Nuclear One Unit 2 109 Replacement Steam Generator Primary Side Channel Head Assy, Sheet 3, Revision 0.

9.24 Drawing No. M-2001-C6-204, Arkansas Nuclear One Unit 2 109 Replacement Steam Generator Upper Shell Assembly, Sheet 2, Revision 0.

9.25 Letter from J. G. Lamb (NRC) to C. A. Gayheart (Southern Nuclear), Vogtle Electric Generating Plant, Units 1 and 2 - Audit Plan for Relief Request Inservice Inspection Alternative VEGP-ISI-ALT-04-04 (EPID L-2020-LLR-0109), ADAMS Accession No. ML20128J311, dated May 14, 2020.

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 (Southern Nuclear), 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 (EPID L-2020-LLR-0109), ADAMS Accession No. ML20258A002, dated December 10, 2020.

0CAN062402 Enclosure Page 24 of 24 9.28 Letter RS-22-084 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, ADAMS Accession No. ML22168A005, dated June 17, 2022.

9.29 Electronic mail 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 (EPID Nos.: L-2021-LLR-091, L-2021-LLR-092, L-2021-LLR-093, L-2021-LLR-094), dated May 6, 2022.

9.30 Drawing No. M-2001-C6-245, Feedwater Nozzle/Thermal Sleeve and Feedwater Ring Assembly, Sheet 2, Revision N.

9.31 UT Vessel Examination Report No. 1-ISI-UT-11-016 (Summary No. 1-03-117), 10/30/11.

9.32 Drawing No. M-2001-C6-204, Arkansas Nuclear One Unit 2 109 Replacement Steam Generator Upper Shell Assembly, Sheet 2, Revision 0.

9.33 Drawing No. M-2001-C6-200, Arkansas Nuclear One Unit 2 109 Replacement Steam Generator Outline, Sheet 2, Revision 0.

9.34 Drawing No. 5022694E (M1D-212), ANO-1 EOTSG Steam Outlet Nozzle Elbows, Revision 3.

9.35 UT Vessel Examination Report No. ISI-UT-08-126 (Summary No. 1-03-118), 11/12/08.

9.36 Entergy Drawing No. MID-295, Sheet 1, Replacement Steam Generator for Arkansas Plant - Unit 1, Principal Side Weld Map Drawing, Repere Des Soudures oe Lenceinte Principale, Revision 1.

9.37 Letter from John G. Lamb (NRC) to Cheryl A. Gayfield (Southern Nuclear Operating Co.,

Inc.), " Vogtle Electric Generating Plant, Units 1 and 2 - Audit Report for the PROMISE Version 1.0 Probabilistic Fracture Mechanics Software used in Relief Request VEGP-ISI-ALT-04-04 (EPID L-2019-LLR-0109)," ADAMS Accession No. ML20128J311, dated December 10, 2020.

ENCLOSURE, ATTACHMENT 1 0CAN062402 PLANT-SPECIFIC APPLICABILITY ANO-1

0CAN062402 Enclosure, Attachment 1 Page 1 of 14 Plant-Specific Applicability - ANO-1 Section 9 of References [1-1] and [1-2] provide requirements that must be demonstrated to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for ANO-1 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 ANO-1.

Table 1-1 Applicability of References [1-1] and [1-2] Representative Analyses to ANO-1 Items No. B2.31 and B2.40 (SG Primary Side Shell Welds)

Category Requirement from Reference [1-1]

Applicability to ANO-1 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. If 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.

The ANO-1 Enhanced Once-Through Steam Generators (EOTSG) 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 a blackout, resulting in thermal shock of any portion of the vessel.

(Notes: Attachment 8.1 of Reference

[1-6] records 3 occurrences of Transient 12 (Hot, Dry, Depressurized SG Refill) in the ANO-1 EOTSG (on 12/20/08, 2/5/09 and 2/7/09).

However, per Attachment 8.2 of Reference [1-6], the actual scenario for each of these 3 occurrences was After manual RX trip, MFW to OTSG temperature dropped below design limit of 135F. CR-ANO-1-2009-00735 conservatively determined this was a Transient 12 cycle. Therefore, the actual transients experienced do not resemble the introducing unheated AFW into a hot, dry SG transient described above. The updated transient counts record no additional occurrences of Transient 12.)

(Reference [1-8]

0CAN062402 Enclosure, Attachment 1 Page 2 of 14 The materials of the SG vessel head, and tube sheet must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The ANO-1 SG vessel heads and tube sheets are fabricated from SA-508 Class 3A material (per Table 3-3 of Reference [1-3]). The maximum Reference Temperature for Nil Ductility Transition (RTNDT) value for the material is 0°F [1-4] and is therefore bounded by the value used in the EPRI report.

SA-508, Class 3A material 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 ANO-1 tube sheet-to-head weld configuration is shown in Figure 1-2 and shows 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 Enclosure, the ANO-1 SG vessel dimensions are as follows:

SG Lower Head Outside Diameter (OD) =

135.00 inches SG Upper Shell OD =

150.75 inches The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [1-1] for Babcock & Wilcox (B&W) 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 ANO-1 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].

0CAN062402 Enclosure, Attachment 1 Page 3 of 14 Items No. C1.30 (SG Secondary Side Shell Welds)

Category Requirement from Reference [1-1]

Applicability to ANO-1 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.

If 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.

The ANO-1 EOTSGs 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 a blackout, resulting in thermal shock of any portion of the vessel.

(Notes: Attachment 8.1 of Reference

[1-6] records 3 occurrences of Transient 12 (Hot, Dry, Depressurized SG Refill) in the ANO-1 EOTSG (on 12/20/08, 2/5/09 and 2/7/09).

However, per Attachment 8.2 of Reference [1-6], the actual scenario for each of these 3 occurrences was After manual RX trip, MFW to OTSG temperature dropped below design limit of 135F. CR-ANO-1-2009-00735 conservatively determined this was a Transient 12 cycle. Therefore, the actual transients experienced do not resemble the introducing unheated AFW into a hot, dry SG transient described above. The updated transient counts record no additional occurrences of Transient 12.)

(Reference [1-8]

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

The ANO-1 SG vessel shell and tube sheet are fabricated from SA-508 Class 3A material (per Table 3-3 of Reference [1-3]). The maximum RTNDT value for the material is 0°F [1-4] and is therefore bounded by the value used in the EPRI report.

SA-508, Class 3A material 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 ANO-1 weld configuration is shown in Figure 1-3 and conforms to Figure 1-8 of Reference [1-1].

0CAN062402 Enclosure, Attachment 1 Page 4 of 14 Category Requirement from Reference [1-1]

Applicability to ANO-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 [1-1].

Per Table 1 of the Enclosure, the ANO-1 SG vessel dimensions are as follows:

SG Lower Head OD =

135.00 inches SG Upper Shell OD =

150.75 inches The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [1-1] for B&W 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 Table 1-3, the ANO-1 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-9 of Reference [1-1].

0CAN062402 Enclosure, Attachment 1 Page 5 of 14 Items Nos. C2.21 and C2.22 (MS and FW Nozzle to Shell Welds and Inside Radius Sections)

Category Requirement from Reference [1-2]

Applicability to ANO-1 General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [1-2].

The ANO-1 MS nozzle-to-shell weld is shown in Figure 1-4 and is representative of the configuration shown in Figure 1-2 of Reference [1-2].

Per Section 4.3.1.3, Item 3 of Reference [1-2], B&W plants (like ANO-1) do not have FW nozzles welded into the SG shells (the nozzle is a bolted joint) and have multiple penetrations in the shell that riser pipes enter to provide feedwater flow to the feedwater ring inside the SG.

There are therefore no C2.21 or C2.22 components for the FW nozzle.

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

The ANO-1 SG vessel shell and MS nozzles are fabricated from SA-508 Class 3A material (per Table 3-3 of Reference [1-3]). The maximum RTNDT value for the materials is 0°F [1-4] and is therefore bounded by the value used in the EPRI report.

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

The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [1-2] over a 60-year operating life.

As shown in Table 1-4, the ANO-1 SGs are not projected to experience more than the number of transients shown in Table 5-5 of Reference [1-2]

over a 60-year operating life.

SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch Nominal Pipe Size (NPS).

There are no C2.21 or C2.22 components for the FW nozzle.

The FW nozzle design must have an integrally attached thermal sleeve.

There are no C2.21 or C2.22 components for the FW nozzle.

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

N/A for ANO-1 (B&W design).

0CAN062402 Enclosure, Attachment 1 Page 6 of 14 Category Requirement from Reference [1-2]

Applicability to ANO-1 SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch NPS.

N/A for ANO-1 (B&W design).

For B&W SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS.

The piping attached to the ANO-1 MS nozzle is 24 inch Inside Diameter (ID)

(per Table 4-4 of Reference [1-3]).

The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG.

The ANO-1 (B&W) MS nozzles are shown in Figure 1-1 and exit the side of the SG.

The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference [1-2]) or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference

[1-2]).

The ANO-1 (B&W) MS nozzles are shown in Figure 1-1 and do not significantly protrude or have a unique nozzle weld configuration.

0CAN062402 Enclosure, Attachment 1 Page 7 of 14 Table 1-2 ANO-1 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head 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]

ANO-1 60-Year Projection Heat up / Cooldown 300 192 / 187(2)

Plant Loading / Unloading 5,000 673 / 673(3)

Reactor Trip 360 187(2)

Notes:

1. Table 5-7 of Reference [1-1] also includes allowable transient temperatures and pressures.

From previous experience with B&W plants, these values are typically within 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. Calculated based on values in Attachment 8.1, PDF pages 5 and 6 (of 45) of Reference [1-6].

It has been confirmed that these calculated values bound those obtained by using Reference [1-8].

3. Table 6 of Reference [1-7].

0CAN062402 Enclosure, Attachment 1 Page 8 of 14 Table 1-3 ANO-1 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel 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]

ANO-1 60-Year Projection Heat up / Cooldown 300 192 / 187(2)

Plant Loading / Unloading 5000 673 / 673(3)

Reactor Trip 360 187(2)

Notes:

1. Table 5-9 of Reference [1-1] also includes allowable transient temperatures and pressures.

From previous experience with B&W plants, these values are typically within 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. Calculated based on values in Attachment 8.1, PDF pages 5 and 6 (of 45) of Reference [1-6]. It has been confirmed that these calculated values bound those obtained by using Reference [1-8].

3. Table 6 of Reference [1-7].

0CAN062402 Enclosure, Attachment 1 Page 9 of 14 Table 1-4 ANO-1 Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles (Comparison to Table 5-5 of Reference [1-2])

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

ANO-1 60-Year Projection Heat up / Cooldown 300 192 / 187(1)

Plant Loading 5000 673(3)

Plant Unloading 5000 673(3)

Loss of Load 360 187(1)(2)

Loss of Power 60 7(4)

Notes:

1. Calculated based on values in Attachment 8.1, PDF pages 5 and 6 (of 45) of Reference [1-6]. It has been confirmed that these calculated values bound those obtained by using Reference [1-8].

2. Loss of Load = Reactor Trip.

3. Table 6 of Reference [1-7].

4. Based on a review of the NRC Licensee Event Report (LER) database for Loss of Offsite Power events at ANO-1. The tally does not factor in the SG replacement and is therefore conservative.

0CAN062402 Enclosure, Attachment 1 Page 10 of 14 Table 1-5 ANO-1 Inspection History Item No.

Component ID Exam Date Interval/Period/Outage Exam Results Coverage Relief Request B2.40 03-103 10/23/2011 4th / 2nd / 1R23 No Recordable Indications (NRI) 100%

N/A B2.40 03-103 5/7/2024 5th / 2nd / 1R31 NRI 100%

N/A B2.40 03-109 4/8/2010 4th / 1st / 1R22 NRI

>90%

N/A B2.40 03-109 10/14/2019 5th / 1st / 1R28 NRI

>90%

N/A B2.31 03-102 10/22/2011 4th / 2nd / 1R23 NRI 73.75%

N/A B2.31 03-102 5/7/2024 5th / 2nd / 1R31 NRI 76%

N/A B2.31 03-110 10/29/2011 4th / 2nd / 1R23 NRI 100%

N/A B2.31 03-110 5/3/2024 5th / 2nd / 1R31 NRI 100%

N/A C1.30 03-104 10/24/2011 4th / 2nd / 1R23 NRI 96.17%

N/A C1.30 03-108 4/6/2010 4th / 1st / 1R22 NRI

>90%

N/A C1.30 03-108 4/12/2018 5th / 1st / 1R27 NRI 99.78%

N/A C2.21 03-117 10/25/2011 4th / 2nd / 1R23 NRI 92.10%

N/A C2.21 03-118 11/9/2008 4th / 1st / 1R21 NRI 92.10%

N/A C2.22 03-115IR 10/26/2011 4th / 2nd / 1R23 NRI 96.76%

N/A C2.22 03-116IR 4/11/2018 5th / 1st / 1R27 NRI 96.76%

N/A

0CAN062402 Enclosure, Attachment 1 Page 11 of 14 Figure 1-1. ANO-1 Steam Generator Layout [1-5]

0CAN062402 Enclosure, Attachment 1 Page 12 of 14 Figure 1-2. ANO-1 Item No. B2.40 Weld Configuration [1-5]

Figure 1-3. ANO-1 Item No. C1.30 Weld Configuration [1-5]

0CAN062402 Enclosure, Attachment 1 Page 13 of 14 Figure 1-4. ANO-1 Main Steam Nozzle Configuration [1-5]

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, Tube sheet-to-Head and Tube sheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

1-2.

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.

1-3.

Engineering Report No. ER-ANO-2002-1381-000, "ANO-1 Steam Generator Replacement," Revision 0.

1-4.

E-mail from Andy Nettles (Entergy) to Scott Chesworth (SI), "

Subject:

RE: ANO Project Inputs," dated May 11, 2023.

1-5.

Drawing No. M1D-295, "Replacement Steam Generator for Arkansas Plant - Unit 1, Principal Side Weld Map Drawing, Repere Des Soudures oe LEnceinte Principale,"

Sheet 1, Revision 1.

1-6.

Entergy Engineering Report No. CALC-ANO1-SE-17-00001, "ANO Unit 1 Transient Cycle Report for 2016," Revision 00.

1-7.

SI Calculation No. 2200654.301, "ANO Units 1 & 2 Plant Loading/Unloading and Pressurizer Insurge/Outsurge Transients," Revision 0.

0CAN062402 Enclosure, Attachment 1 Page 14 of 14 1-8.

Entergy Engineering Report No. CALC-ANO1-SE-18-00002, "ANO Unit 1 Transient Cycle Report for 2017 Through 2022," Revision 00.

ENCLOSURE, ATTACHMENT 2 0CAN062402 PLANT-SPECIFIC APPLICABILITY ANO-2

0CAN062402 Enclosure, Attachment 2 Page 1 of 16 Plant-Specific Applicability - ANO-2 Section 9 of References [2-1] and [2-2] provide requirements that must be demonstrated to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for ANO-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 ANO-2.

Table 2-1 Applicability of References [2-1] and [2-2] Representative Analyses to ANO-2 Items No. B2.40 (Steam Generator (SG) Primary Side Shell Welds)

Category Requirement from Reference [2-1]

Applicability to ANO-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.

If 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.

The ANO-2 Replacement Steam Generators (RSGs) 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 a blackout, resulting in thermal shock of any portion of the vessel.

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

The ANO-2 RSG vessel heads and tube sheet are fabricated from SA-508 Class 3A material (per Table 5.2-3 of Reference [2-3]). The Reference Temperature for Nil Ductility Transition (RTNDT) value for the material is assumed to be 60°F, the maximum value allowed by Branch Technical Position (BTP) 5-3 [2-4] and therefore consistent with that used in the EPRI report.

SA-508, Class 3A material 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 [2-1].

The ANO-2 tube sheet-to-head weld configuration is shown in Figure 2-2 and shows conformance with Figure 1-2 of Reference [2-1].

0CAN062402 Enclosure, Attachment 2 Page 2 of 16 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].

Using the Inside Diameter (ID) and thickness values from Table 1 of the Enclosure, the ANO-2 SG Outside Diameter (OD) vessel dimensions are as follows:

SG Lower Head OD =

171.30 inches SG Upper Shell OD =

240.45 inches The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [2-1] for Combustion Engineering (CE) plants.

(Note: The values given in Table 5.5-2 of Reference [2-3] are 171.07 inches and 240.69 inches respectively; these values are extremely close to the values above from Table 1.)

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 ANO-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].

0CAN062402 Enclosure, Attachment 2 Page 3 of 16 Item No. B3.130 (SG Primary Inlet/Outlet Nozzles)

Category Requirement from Reference [2-1]

Applicability to ANO-2 General Requirements The Loss of Power transient (involving unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. If such a significant thermal event occurs at a plant, its impact on the KIC value may require more frequent examinations and other plant actions outside the scope of this reports guidance.

The ANO-2 RSGs 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 a blackout, resulting in thermal shock of any portion of the vessel.

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

The ANO-2 RSG vessel heads and tube sheet are fabricated from SA-508 Class 3A material (per Table 5.2-3 of Reference [2-3]). The RTNDT value for the material is assumed to be 60°F, the maximum value allowed by BTP 5-3 [2-4] and therefore consistent with that used in the EPRI report.

SA-508, Class 3A material 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 1-3 through 1-5 of Reference [2-1].

The ANO-2 weld configurations are shown in Figure 2-3 and conform to those shown in Figures 1-3 through 1-5 of Reference [2-1].

The piping attached to the primary inlet and outlet nozzles (RCS piping) for the various designs must be within 10% of the values provided in the table in Section 9.4.2 of Reference [2-1].

The piping attached to the ANO-2 primary inlet and outlet nozzles (RCS piping) is 42 inches ID and 30 inches ID, respectively (per Table 5.5-2 of

[2-3]). These values are within 10% of the values provided in the table in Section 9.4.2 of Reference [2-1].

The component must experience transients and cycles bounded by those shown in Table 5-8 of Reference [2-1] over a 60-year operating life.

As shown in Table 2-3, the ANO-2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-8 of Reference [2-1].

0CAN062402 Enclosure, Attachment 2 Page 4 of 16 Items No. C1.20 and C1.30 (SG Secondary Side Shell Welds)

Category Requirement from Reference [2-1]

Applicability to ANO-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.

If 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.

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

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

The ANO-2 RSG vessel heads and tube sheet are fabricated from SA-508 Class 3A material (per Table 5.2-3 of Reference [2-3]). The RTNDT value for the material is assumed to be 60°F, the maximum value allowed by BTP 5-3 [2-4] and therefore consistent with that used in the EPRI report.

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

0CAN062402 Enclosure, Attachment 2 Page 5 of 16 Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [2-1].

The ANO-2 weld configurations are shown in Figures 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].

Using the ID and thickness values from Table 1 in the Enclosure, the ANO-2 SG OD vessel dimensions are as follows:

SG Lower Head OD =

171.30 inches SG Upper Shell OD =

240.45 inches The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [2-1] for CE plants.

(Note: The values given in Table 5.5-2 of Reference [2-3] are 171.07 inches and 240.69 inches respectively; these values are extremely close to the values above from Table 1.)

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 2-4, the ANO-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].

0CAN062402 Enclosure, Attachment 2 Page 6 of 16 Items Nos. C2.21 and C2.22 (Main Steam (MS) and Feedwater (FW) Nozzle to Shell Welds and Inside Radius Sections)

Category Requirement from Reference [2-2]

Applicability to ANO-2 General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [2-2].

The ANO-2 FW nozzle-to-shell weld is shown in Figure 2-6 and is representative of the configuration shown in Figure 1-2 of Reference [2-2].

The ANO-2 MS nozzle-to-shell weld is shown in Figure 2-7 and is representative of the configuration shown in Figure 1-2 of Reference [2-2]

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

The ANO-2 RSG shell, FW nozzles, and MS nozzles are fabricated from SA-508 Class 3A material (per Table 5.2-3 of Reference [2-3]). The RTNDT value for the material is assumed to be 60°F, the maximum value allowed by BTP 5-3 [2-4] and therefore consistent with that used in the EPRI report.

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

The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [2-2] over a 60-year operating life.

As shown in Table 2-5, the ANO-2 SGs are not projected to experience more than the number of transients shown in Table 5-5 of Reference [2-2]

over a 60-year operating life.

SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch Nominal Pipe Size (NPS).

The piping attached to the ANO-2 FW nozzle is 18 inches nominal size (per Table 5.5-2 of [2-3]).

The FW nozzle design must have an integrally attached thermal sleeve.

The ANO-2 FW nozzle design has an integrally attached thermal sleeve per Reference [2-8].

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

N/A for ANO-2 (CE Design).

SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch Nominal Pipe Size (NPS).

The piping attached to the ANO-2 (CE) main steam nozzle is 34 ID (per Table 5.5-2 of [2-3]).

0CAN062402 Enclosure, Attachment 2 Page 7 of 16 Category Requirement from Reference [2-2]

Applicability to ANO-2 For Babcock and Wilcox (B&W) SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS.

N/A for ANO-2 (CE design).

The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG.

The ANO-2 (CE) MS nozzle is shown in Figure 2-1 and exits the top dome of the SG.

The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference [2-2]) or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference

[2-2]).

The ANO-2 MS nozzle is shown in Figure 2-7 and does not significantly protrude or have a unique nozzle configuration.

Table 2-2 ANO-2 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head 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]

ANO-2 60-Year Projection Heat up / Cooldown 300 237 / 237(2)

Plant Loading /

Unloading 5000 493 / 493(3)

Reactor Trip 360 156(2)

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. Calculated based on values on Page 8 (of 46) of Reference [2-11].

3. Table 7 of Reference [2-12].

0CAN062402 Enclosure, Attachment 2 Page 8 of 16 Table 2-3 ANO-2 Data for Thermal Transients for Stress Analysis of the PWR SG Inlet Nozzle-to-Vessel Welds (Item No. B3.130) (Comparison to Table 5-8 of Reference [2-1])

Transient(1)

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

ANO-2 60-Year Projection Heat up / Cooldown 300 237 / 237(2)

Plant Loading /

Unloading 5000 493 / 493(3)

Reactor Trip 360 156(2)

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. Calculated based on values on Page 8 (of 46) of Reference [2-11].

3. Table 7 of Reference [2-12].

0CAN062402 Enclosure, Attachment 2 Page 9 of 16 Table 2-4 ANO-2 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel 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]

ANO-2 60-Year Projection Heat up / Cooldown 300 237 / 237(2)

Plant Loading /

Unloading 5000 493 / 493(3)

Reactor Trip 360 156(2)

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. Calculated based on values on Page 8 (of 46) of Reference [2-11].

3. Table 7 of Reference [2-12].

0CAN062402 Enclosure, Attachment 2 Page 10 of 16 Table 2-5 ANO-2 Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles (Comparison to Table 5-5 of Reference [2-2])

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

ANO-2 60-Year Projection Heat up / Cooldown 300 237 / 237(1)

Plant Loading 5000 493(3)

Plant Unloading 5000 493(3)

Loss of Load 360 156(1)(2)

Loss of Power 60 9(4)

Notes:

1. Calculated based on values on Page 8 (of 46) of Reference [2-11].

2. Loss of Load = Reactor Trip.

3. Table 7 of Reference [2-12].

4. Based on a review of the NRC Licensee Event Report (LER) database for Loss of Offsite Power events at ANO-2.

0CAN062402 Enclosure, Attachment 2 Page 11 of 16 Table 2-6 ANO-2 Inspection History Item No.

Component ID Exam Date Interval/Period/Outage Exam Results Coverage Relief Request B2.40 04-004 10/10/2006 3rd / 2nd / 2R18 No Recordable Indications (NRI)

>90%

N/A B2.40 04-004 10/17/2015 4th / 2nd / 2R24 NRI 100%

N/A B3.130 03-005 5/19/2014 4th / 2nd / 2R23 NRI 87%

N/A B3.130 04-005 10/8/2006 3rd / 2nd / 2R18 NRI

<90%

N/A B3.130 04-005 10/3/2015 4th / 2nd / 2R24 NRI 88%

N/A B3.130 03-006 5/19/2014 4th / 2nd / 2R23 NRI 87%

N/A B3.130 04-006 10/8/2006 3rd / 2nd / 2R18 NRI

<90%

N/A B3.130 04-006 10/3/2015 4th / 2nd / 2R24 NRI 88%

N/A B3.130 03-007 5/30/2014 4th / 2nd / 2R23 NRI 94%

N/A B3.130 04-007 10/10/2006 3rd / 2nd / 2R18 NRI

<90%

N/A B3.130 04-007 10/10/2015 4th / 2nd / 2R24 NRI 87%

N/A C1.20 04-001 10/8/2015 4th / 2nd / 2R24 NRI 100%

N/A C1.30 04-003 10/16/2006 3rd / 2nd / 2R18 NRI

>90%

N/A C1.30 04-003 10/12/2018 4th / 3rd / 2R26 NRI 98.90%

N/A C2.21 04-002 10/12/2006 3rd / 2nd / 2R18 NRI

<90%

N/A C2.21 04-002 10/7/2015 4th / 2nd / 2R24 NRI 69.60%

N/A C2.22 04-002IR 10/7/2015 4th / 2nd / 2R24 NRI 100%

N/A

0CAN062402 Enclosure, Attachment 2 Page 12 of 16 Figure 2-1. ANO-2 Steam Generator Layout [2-5]

0CAN062402 Enclosure, Attachment 2 Page 13 of 16 Figure 2-2. ANO-2 Item No. B2.40 Weld Configuration [2-6]

Figure 2-3. ANO-2 Item No. B3.130 Weld Configuration [2-6]

0CAN062402 Enclosure, Attachment 2 Page 14 of 16 Figure 2-4. ANO-2 Item No. C1.20 Weld Configuration [2-10]

Figure 2-5. ANO-2 Item No. C1.30 Weld Configuration [2-7]

0CAN062402 Enclosure, Attachment 2 Page 15 of 16 Figure 2-6. ANO-2 Feedwater Nozzle Configuration [2-8]

Figure 2-7. ANO-2 Main Steam Nozzle Configuration [2-9]

0CAN062402 Enclosure, Attachment 2 Page 16 of 16 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, Tube sheet-to-Head and Tube sheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

2-2.

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.

2-3.

Arkansas Nuclear One - Unit 2, SAR Amendment 22, Facility Operating License Number NPF-6, Docket Number 50-368.

2-4.

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

2-5.

Drawing No. M-2001-C6-200, "Arkansas Nuclear One Unit 2 109 Replacement Steam Generator Outline," Sheet 2, Revision 0.

2-6.

Drawing No. M-2001-06-293, "Arkansas Nuclear One Unit 2 109 Replacement Steam Generator, Primary Side Channel Head Assy," Sheet 2, Revision 0.

2-7.

Drawing No. M-2001-C6-213, "Steam Generator Tube Plate and Lower Barrel Weld Assembly and Machining," Sheet 1, Revision 0.

2-8.

Drawing No. M-2001-C6-245, "Feedwater Nozzle/Thermal Sleeve and Feedwater Ring Assembly," Sheet 2, Revision N.

2-9.

Drawing No. M-2001-C6-204, "Arkansas Nuclear One Unit 2 109 Replacement Steam Generator, Upper Shell Assembly," Sheet 2, Revision 0.

2-10. Drawing No. M-2001-C6-204, "Arkansas Nuclear One Unit 2 109 Replacement Steam Generator, Upper Shell Assembly," Sheet 1, Revision 0.

2-11. Entergy Engineering Report No. CALC-ANO2-SE-18-00002, "ANO Unit 2 Transient Cycle Report for 2018," Revision 0.

2-12. SI Calculation No. 2200654.301, "ANO Units 1 & 2 Plant Loading/Unloading and Pressurizer Insurge/Outsurge Transients," Revision 0.

ENCLOSURE, ATTACHMENT 3 0CAN062402 RESULTS OF INDUSTRY SURVEY

0CAN062402 Enclosure, Attachment 3 Page 1 of 4 Results of Industry Survey 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 Steam Generator (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 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 American Society of Mechanical Engineers (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 per IWC-3600 was performed for these flaws and they were found to be acceptable for continued operation.

0CAN062402 Enclosure, Attachment 3 Page 2 of 4 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 Westinghouse (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 [3-4, 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.

0CAN062402 Enclosure, Attachment 3 Page 3 of 4 Overall Industry Inspection Summary for Code Items C2.21, C2.22, and C2.32 The results of an industry survey of past inspections of SG main steam (MS) and feedwater (FW) nozzles are summarized in Reference [3-2]. Table 3-2 provides a summary of the combined survey results for Item Nos. C2.22, C2.21, and C2.32 (see Table Note 1). The results identify that SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Section 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 BWR and 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 NSSS vendors (i.e., B&W, CE, and Westinghouse). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32 (see Table Note 1) components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG MS and FW nozzles. Only one PWR examination identified two (2) flaws that exceeded ASME Code,Section XI acceptance criteria. The flaws were linear indications of 0.3 and 0.5 in length and were detected in a MS nozzle-to-shell weld using magnetic particle examination techniques. The indications were dispositioned by light grinding (ADAMS Accession No. ML13217A093).

Table 3-2 Summary of Survey Results for SG Main Steam and Feedwater Nozzle Components Plant Type Number of Units Number of Examinations Number of Reportable Indications BWR 27 164 0

PWR 47 563 2

Totals 74 727 (Note 1) 2 Notes:

1. Item No. C2.32 was evaluated in the Reference [3-2] technical basis and included in the industry survey but is not contained in the scope of this alternative request.

0CAN062402 Enclosure, Attachment 3 Page 4 of 4 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, Tube sheet-to-Head and Tube sheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

3-2.

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.

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)," ADAMS Accession Number ML13217A093, dated July 29, 2013.

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," ADAMS Accession Number ML011550346, dated May 29, 2001.

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' (Tac Nos. MB7294 and MB7295)," ADAMS Accession Number ML031040553, dated April 4, 2003.