L-24-233, Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examinations for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles

Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examinations for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles

ML24353A315
Person / Time
Site:	Davis Besse
Issue date:	12/18/2024
From:	Tony Brown Vistra Operations Company
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-24-233
Download: ML24353A315 (1)

Text

Davis-Besse Nuclear Power Station Terry J. Brown Site Vice President 5501 N. State Route 2 Oak Harbor, Ohio 43449 419.321.7676 L-24-233 December 18, 2024 10 CFR 50.55a(z)(1)

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examinations for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles In accordance with 10 CFR 50.55a, Codes and Standards, paragraph (z)(1), Vistra Operations Company LLC (Vistra OpCo) hereby requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative to the inservice inspection (ISI) requirements for American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, Table IWB-2500-1, Examination Category B-B and B-D and Table IWC-2500-1, Examination Category C-A and C-B, component examinations for Davis-Besse Nuclear Power Station, Unit No. 1, (Davis-Besse). Specifically, Vistra OpCo requests an increase to the inspection interval for the applicable examination items from the current ASME Code,Section XI 10-year requirement, thereby deferring examinations for two 10-year ISI intervals from the last examination performed for each item. The subject welds will be reexamined prior to the end of the current 60-year operating license for Davis-Besse, which expires on April 22, 2037. The proposed alternative is requested on the basis that it provides an acceptable level of quality and safety in lieu of the current ASME Code,Section XI 10-year inspection frequency requirement.

The proposed alternative request RR-A1, which includes a summary of the technical basis for the request, is provided in Attachment 1. The plant-specific applicability of the technical basis to the Davis-Besse steam generators and pressurizer are provided in Attachments 2 and 3, respectively. An industry inspection survey summary for the applicable component code items is presented in Attachment 4. Deterministic fracture mechanics and probabilistic fracture mechanics evaluations performed by Structural Integrity Associates, Inc. (SI) for the Babcock and Wilcox design pressurizers are provided for convenience in Attachments 5 and 6.

Vistra requests NRC approval of the proposed alternative request by December 19, 2025, to support the Davis-Besse spring 2026 refueling outage.

Davis-Besse Nuclear Power Station, Unit No. 1 L-24-233 Page2 There are no regulatory commitments contained in this submittal. If there are any questions, or if additional information is required, please contact Jack Hicks, Senior Manager, Fleet Licensing, at (254) 897-6725 or jack.hicks@vistracorp.com.

Attachments:

1.

10 CFR 50.55a Request Number: RR-Al

2.

Plant-Specific Applicability Davis-Besse Steam Generator

3.

Plant-Specific Applicability Davis-Besse Pressurizer

4.

Results of Industry Survey

5.

SI Calculation No. 2100561.302, "Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head," Revision 1

6.

SI Calculation No. 2100561.303, "Deterministic and Probablistic Fracture Mechanics of Oconee Units 1, 2 and 3 Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head," Revision 2 cc:

NRC Region III Administrator NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board L-24-233 10 CFR 50.55a Request Number: RR-A1 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

Page 1 of 26

1.

ASME Code Component(s) Affected Steam Generators 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 2, Category C-A, pressure-retaining welds in pressure vessels Class 2, Category C-B, pressure-retaining nozzle welds in pressure vessels Item Numbers:

B2.40 - Steam generators (primary side), tubesheet-to-head weld C1.30 - Tubesheet-to-shell weld C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections Steam Generator Components ASME Category ASME Item No.

Component ID Component Description B-B B2.40 RC-SG-1-1-W23 Upper Tubesheet to Upper Primary Head Weld B-B B2.40 RC-SG-1-2-W23 Upper Tubesheet to Upper Primary Head Weld B-B B2.40 RC-SG-1-1-W22 Lower Tubesheet to Lower Primary Head Weld B-B B2.40 RC-SG-1-2-W22 Lower Tubesheet to Lower Primary Head Weld C-A C1.30 SP-SG-1-1-W65 Shell to Lower Tubesheet Weld C-A C1.30 SP-SG-1-1-W69 Upper Tubesheet to Shell Weld C-A C1.30 SP-SG-1-2-W65 Shell to Lower Tubesheet Weld C-A C1.30 SP-SG-1-2-W69 Upper Tubesheet to Shell Weld C-B C2.21 SP-SG-1-1-W127-X/Y 24 in. X/Y Axis Steam Outlet Nozzle to Shell Weld C-B C2.21 SP-SG-1-1-W128-W/X 24 in. W/X Axis Steam Outlet Nozzle to Shell Weld C-B C2.21 SP-SG-1-2-W127-X/Y 24 in. X/Y Axis Steam Outlet Nozzle to Shell Weld C-B C2.21 SP-SG-1-2-W128-W/X 24 in. W/X Axis Steam Outlet Nozzle to Shell Weld C-B C2.22 SP-SG-1-1-W127-X/Y-IR 24 in. X/Y Axis Steam Outlet Nozzle Inside Radius C-B C2.22 SP-SG-1-1-W128-W/X-IR 24 in. W/X Axis Steam Outlet Nozzle Inside Radius C-B C2.22 SP-SG-1-2-W127-X/Y-IR 24 in. X/Y Axis Steam Outlet Nozzle Inside Radius C-B C2.22 SP-SG-1-2-W128-W/X-IR 24 in. W/X Axis Steam Outlet Nozzle Inside Radius L-24-233 Page 2 of 26 Pressurizer Code Class:

Class 1

==

Description:==

Pressurizer vessel head, shell-to-head, and nozzle-to-vessel welds Examination Category:

Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 1, Category B-D, full penetration welded nozzles in vessels Item Numbers:

B2.11 - Pressurizer, shell-to-head welds, circumferential B2.12 - Pressurizer, shell-to-head welds, longitudinal B3.110 - Pressurizer, nozzle-to-vessel welds Pressurizer Components ASME Category ASME Item No.

Component ID Component Description B-B B2.11 RC-PZR-WP-76 Shell-to-Head Circumferential Weld B-B B2.11 RC-PZR-WP-28 Shell-to-Head Circumferential Weld B-B B2.12 RC-PZR-WP-1 Shell-to-Head Longitudinal Weld(1)

B-B B2.12 RC-PZR-MK-4-40-6-WP-7-Y-LU Shell-to-Head Longitudinal Weld(1)

B-B B2.12 RC-PZR-MK-4-40-6-WP-7-X-LU Shell-to-Head Longitudinal Weld(1)

B-D B3.110 RC-PZR-WP-33-W/X Nozzle-to-Head Weld B-D B3.110 RC-PZR-WP-34 Nozzle-to-Head Weld B-D B3.110 RC-PZR-WP-33-Z/W Nozzle-to-Head Weld B-D B3.110 RC-PZR-WP-15 Nozzle-to-Head Weld B-D B3.110 RC-PZR-WP-33-Y/Z Nozzle-to-Head Weld Note 1: The applicable portion of the longitudinal seam weld is where it intersects the associated Item No. B2.11 (shell-to-head) weld.

2.

Applicable Code Edition and Addenda

The fifth 10-year inservice inspection (ISI) interval Code of record for Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse) is the 2017 Edition of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

3.

Applicable Code Requirement

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

B2.11 Volumetric examination of both circumferential shell-to-head welds during each inspection interval. The examination volume is shown in Figure IWB-2500-1.

L-24-233 Page 3 of 26 B2.12 Volumetric examination of one foot of all longitudinal shell-to-head welds that intersect circumferential welds during the first interval and one foot of one longitudinal shell-to-head weld per head that intersects a circumferential weld during successive intervals. The examination volume is shown in Figure IWB-2500-2.

B2.40 Volumetric examination of essentially 100 percent 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.

B3.110 Volumetric examination of all full penetration nozzle-to-vessel welds during each inspection interval. The examination volume is shown in Figures IWB-2500-7(a), (b), (c), or (d).

C1.30 Volumetric examination of essentially 100 percent of the weld length of the tubesheet-to-shell welds during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, 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.

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 SGs, 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).

C2.22 Volumetric examination of all nozzle inside radius sections at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers),

the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figures IWC-2500-4(a), (b),

or (d).

4.

Reason for Request

The Electric Power Research Institute (EPRI) performed assessments in References

[9.1], [9.2] and [9.3] 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 and pressurizer 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], [9.2] and [9.3] reports concluded that the current ASME Code,Section XI ISI examinations can be deferred for L-24-233 Page 4 of 26 some time with no impact to plant safety. Based on the conclusions of the three EPRI reports, supplemented by plant-specific evaluations contained herein, Vistra Operations Company LLC (Vistra OpCo) is requesting an ISI examination deferral for the subject welds at Davis-Besse. The Reference [9.1], [9.2] and [9.3] reports were developed consistent with the recommendations provided in EPRIs White Paper on suggested content for PFM submittals [9.4] and Nuclear Regulatory Commission (NRC) Regulatory Guide 1.245 for PFM submittals and associated technical basis [9.5, 9.6].

5.

Proposed Alternative and Basis for Use For Davis-Besse, Vistra OpCo 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 SG categories and item numbers:

ASME Category Item No.

Description B-B B2.40 Steam generators (primary side), tubesheet-to-head welds C-A C1.30 Steam generators (secondary side), tubesheet-to-shell welds C-B C2.21 Steam generators (secondary side), nozzle-to-shell welds C-B C2.22 Steam generators (secondary side) nozzle inside radius sections In 2014 (first period of the fourth inspection interval), both Davis-Besse SGs were replaced. The new SG welds and components received the required preservice inspection (PSI) examinations prior to service followed by ISI examinations through the third period of the fourth inspection interval (Davis-Besse is currently in the first period of the fifth inspection interval).

Vistra OpCo is also requesting an inspection alternative to the examination requirements of ASME Code,Section XI, Table IWB-2500-1 for the following examination pressurizer categories and item numbers.

ASME Category Item No.

Description B-B B2.11 Pressurizers, shell-to-head welds, circumferential B-B B2.12 Pressurizers, shell-to-head welds, longitudinal B-D B3.110 Pressurizers, nozzle-to-vessel welds The proposed alternative is to increase the inspection interval for all of these SG and pressurizer item numbers from the current ASME Code,Section XI 10-year requirement, thereby deferring examination for two 10-year ISI intervals from the last examination performed for each item number. The following tables reflect the proposed examination schedule for the required SG and pressurizer examinations. Steam Generator 1 is the selected vessel to comply with Item Nos. B2.40, C1.30, C2.21, and C2.22 requirements; however, Steam Generator 2 equivalent weld examinations may be substituted to satisfy the requirements.

L-24-233 Page 5 of 26 Steam Generator Item No.

Component ID Last Examination Performed Proposed Examination Schedule B2.40 RCSG11W23 4th Interval / 3rd Period 6th Interval / 2nd Period B2.40 RCSG12W23 4th Interval / 1st Period 6th Interval / 1st Period B2.40 RCSG11W22 4th Interval / 1st Period 6th Interval / 1st Period B2.40 RCSG12W22 4th Interval / 2nd Period 6th Interval / 2nd Period C1.30 SPSG11W65 4th Interval / 1st Period 6th Interval / 1st Period C1.30 SPSG11W69 4th Interval / 3rd Period 6th Interval / 2nd Period C1.30 SPSG12W65 4th Interval / 2nd Period 6th Interval / 2nd Period C1.30 SPSG12W69 4th Interval / 1st Period 6th Interval / 1st Period C2.21 SPSG11W127X/Y 4th Interval / 1st Period 6th Interval / 1st Period C2.21 SPSG11W128W/X 4th Interval / 3rd Period 6th Interval / 2nd Period C2.21 SPSG12W127X/Y 4th Interval / 2nd Period 6th Interval / 2nd Period C2.21 SPSG12W128W/X 4th Interval / 1st Period 6th Interval / 1st Period C2.22 SPSG11W127X/YIR 4th Interval / 1st Period 6th Interval / 1st Period C2.22 SPSG11W128W/X-IR 4th Interval / 3rd Period 6th Interval / 2nd Period C2.22 SPSG12W127X/YIR 4th Interval / 2nd Period 6th Interval / 2nd Period C2.22 SPSG12W128W/X-IR 4th Interval / 1st Period 6th Interval / 1st Period Pressurizer Item No.

Component ID Last Examination Performed Proposed Examination Schedule B2.11 RC-PZR-WP-28 4th Interval / 3rd Period 6th Interval / 2nd Period B2.11 RC-PZR-WP-76 4th Interval / 1st Period 6th Interval / 1st Period B2.12 RC-PZR-MK-4-40-6-WP-7-Y-LU 4th Interval / 3rd Period 6th Interval / 2nd Period B2.12 RC-PZR-WP-1 4th Interval / 1st Period 6th Interval / 1st Period B3.110 RC-PZR-WP-15 4th Interval / 3rd Period 6th Interval / 2nd Period B3.110 RC-PZR-WP-33-W/X 4th Interval / 1st Period 6th Interval / 1st Period B3.110 RC-PZR-WP-33-Y/Z 4th Interval / 3rd Period 6th Interval / 2nd Period B3.110 RC-PZR-WP-33-Z/W 4th Interval / 2nd Period 6th Interval / 2nd Period B3.110 RC-PZR-WP-34 4th Interval / 1st Period 6th Interval / 1st Period Technical Basis A summary of the key aspects of the technical bases for this request is summarized below. The applicability of the technical bases to the Davis-Besse SGs and pressurizer is shown in Attachments 2 and 3, respectively.

L-24-233 Page 6 of 26 Applicability of the Degradation Mechanism Evaluation in References [9.1], [9.2], and [9.3]

to Davis-Besse An evaluation of degradation mechanisms that could potentially impact the reliability of the SG and pressurizer welds and components was performed in References [9.1], [9.2] and

[9.3]. The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC),

pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC),

general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no known active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG and pressurizer 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

[9.7] Safety Evaluation (SE) for Vogtle Units 1 and 2; Section 2.0, page 3, second paragraph of the Reference [9.8] SE for Millstone Unit 2; and Section 2, page 3, second paragraph of the Reference [9.9] SE for Salem Units 1 and 2. As shown in Attachments 2 and 3, the materials and operating conditions for the Davis-Besse SG and pressurizer welds and components considered in this request for alternative are similar to those in the References [9.1], [9.2] and [9.3] and therefore, the conclusions of these reports apply to the components in this request for alternative. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in References [9.1], [9.2] and [9.3].

As part of the technical basis in References [9.1], [9.2] and [9.3], a comprehensive industry survey involving 74 pressurized water reactor (PWR) and boiling water reactor (BWR) 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], [9.2] and [9.3] to Davis-Besse Steam Generators 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 analyses in References [9.1] and [9.2] were 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 the Davis-Besse SGs is demonstrated in Attachment 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 Davis-Besse in Tables 1 and 2:

L-24-233 Page 7 of 26 Table 1. SG Vessel Dimensions Plant Primary Lower Head ID (in)

Primary Lower Head Thk (in)

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

Secondary Upper Shell Thk (in)

Secondary Upper Shell Ri/t EPRI Report (Table 4-2 of [9.2])

151.24 6.94 10.9 230.87 4.91 23.5 Davis-Besse 119.06 [9.32]

6.375 [9.32]

9.34 137.875 [9.32]

3.125 [9.32]

22.06 Table 2. SG Nozzle Dimensions Plant FW Nzl ID (in)

FW Nzl Thk (in)

FW Nzl Ri/t MS Nzl ID (in)

MS Nzl Thk (in)

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

16.5 6

1.38 22.25 4.53 2.46 Davis-Besse N/A N/A N/A 20.38 [9.33]

3.5 [9.32]

2.91 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 [9.7], 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 ratios (Ri/t) of the Davis-Besse SG welds and components relative to those used in the EPRI reports are as follows.

Primary lower head: (9.34/11.2) = 0.834 (applicable to primary side welds but conservatively assumed applicable to the rest of the SG welds)

Secondary upper shell: (22.06/23.5) = 0.939 (applicable to main steam [MS] nozzle-to-shell welds)

Feedwater (FW) nozzle: N/A for Davis-Besse MS nozzle: (2.91/2.46) = 1.18 (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 Davis-Besse are performed at normal operating conditions. No hydrostatic testing has been performed at Davis-Besse since the SGs went into operation.

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 [9.10], RAI 3c), the NRC raised this issue. In response to the RAI (Reference [9.11], RAI Response 3.c), an evaluation was performed that 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 [9.8].

L-24-233 Page 8 of 26 Pressurizer The geometric configuration of the pressurizer used in the Reference [9.3] stress analysis, while consistent with the Westinghouse/Combustion Engineering (CE) pressurizer designs, is not appropriate for the Babcock and Wilcox (B&W) pressurizer design at Davis-Besse. A plant-specific stress analysis was previously performed for the B&W pressurizers at Oconee Nuclear Station Units 1, 2, and 3 (ONS1/2/3) and is documented in Attachment 7 of Reference [9.12]. For convenience, Attachment 7 of Reference [9.12] is attached as Attachment 5 of this request for alternative. Table 3 (below) provides a comparison of key plant-specific parameters for the Davis-Besse and ONS1/2/3 pressurizers and demonstrates consistency between the B&W pressurizers at the two plants. Therefore, the plant-specific stress analysis performed for ONS1/2/3 in, together with additional information provided in Attachment 3, demonstrates that all plant-specific requirements are met for Davis-Besse.

Table 3. Comparison of Key Parameters of B&W Pressurizers at Davis-Besse and ONS1/2/3 Parameter ONS1/2/3 Value Davis-Besse Value Shell/head material SA-516 Grade 70(1)(2)

SA-516 Grade 70 [9.34]

Nozzle material SA-508 Grade 1 Class 1(1)

A508 Class 1 [9.34]

Shell ID (in) 84(3) 84(4)

Shell Thk away from heaters (in) 6.188(3) 6.188(5)

Shell Thk near heaters (in) 13.5625(3) 13.5625 [9.37]

Surge Nzl ID (in) 9.25(3) 9.25 [9.38]

Surge Nzl Thk (in) 3(3) 3(6)

Notes:

1.

Section 5.1.1 of Attachment 7 of Reference [9.12].

2.

ONS2/3 fabricated of SA-516 Grade 70 carbon steel. ONS1 fabricated of SA-212 Grade B which is a predecessor to SA-516 Grade 70 and has similar material properties.

3.

Figure 1 of Attachment 7 of Reference [9.12].

4.

Calculated from OD and minimum thickness values in Table 5.1-4 of Reference [9.35].

5.

Table 5.1-4 of Reference [9.35].

6.

Calculated from OD and ID values of Reference [9.38].

The technical approach used in the stress analysis for the B&W pressurizer design in is consistent with Section 7 of the Reference [9.3] report, using the ONS1/2/3 plant-specific geometry and operating conditions. Based on the results in the Reference [9.3] EPRI report, the bottom head is controlling from a stress point of view due to the insurge/outsurge transients experienced in that region. Hence, the plant-specific stress analyses for the B&W pressurizer design were performed for the bottom head. The stress results are presented in Section 6.0 of Attachment 5. Because of the relatively complicated geometry of the B&W pressurizer design (shown in Figure 1 of ), thirteen (13) critical stress paths were chosen for subsequent fracture mechanics evaluations, compared to two in the Reference [9.3] EPRI report. The L-24-233 Page 9 of 26 locations of the 13 critical stress paths are provided in Figure 13 of Attachment 5, which corresponds to Figure 7-9 of the Reference [9.3] EPRI Report. Typical transient stresses for the 13 stress paths are provided in Figures 14 through 26 of Attachment 7 of Reference [9.12], which correspond to Figures 7-10 and 7-11 of the Reference [9.3] EPRI report.

Applicability of the Flaw Tolerance Evaluation in References [9.1], [9.2] and [9.3] to Davis-Besse Steam Generators 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 ISIs, the NRCs safety goal of 1.0x10-6 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 an alternative request submitted by Southern Nuclear Company (SNC), the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan dated May 14, 2020 (ADAMS Accession No. ML20128J311) and the audit summary report issued by letter dated December 10, 2020 (ADAMS Accession No. ML20258A002). 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 percent 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 [9.7], 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 1.0x10-6 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 Davis-Besse is provided below. The assumption below of a 30-year ISI deferral is conservative compared to the end of currently licensed operating life for each plant.

For the Davis-Besse replacement SGs installed in 2014 (first period of the fourth inspection interval), PSI examinations have been performed followed by ISI examinations over one completed 10-year interval following SG replacement. The PSI/ISI scenario considered is therefore PSI plus one set of 10-year ISI examinations to be followed by two 30-year ISI deferrals (PSI+10+40+70).

L-24-233 Page 10 of 26 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 Davis-Besse B&W SG FW nozzle configuration; therefore, the MS nozzles are considered at Davis-Besse. From Reference [9.1], the critical Case ID for the main steam nozzle inside radius section is SGB-P1N. An evaluation similar to that shown in Table 8-28 of Reference [9.1] was performed for this location assuming a nozzle flaw density of 0.1, a stress multiplier of 1.5, a fracture toughness of 200 ksiin and a standard deviation 5 ksiin as recommended by the NRC in Reference [9.7]. A relatively high stress multiplier of 2.35 was used, which represents the highest stress multiplier that can be applied without exceeding the acceptance criterion of 1.0x10-6. The results of the evaluation are summarized in Table 4 up to 80 years of plant operation for the PSI/ISI inspection scenario of (PSI+10+40+70).

Table 4. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the Davis-Besse Main Steam Nozzle Inside Radius Section (Case ID SGB-P1N from Reference [9.1])

Time (year)

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 For the main steam nozzle-to-shell weld, Table 8-15 of Reference [9.1] indicates that the critical Case ID is SGB-P3A. For the current evaluation, a nozzle 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 used, which represents the highest stress multiplier that can be applied without exceeding the acceptance criterion of 1.0x10-6. The results of the evaluation are summarized in Table 5 up to 80 years of plant operation for the PSI/ISI inspection scenario of (PSI+10+40+70).

L-24-233 Page 11 of 26 Table 5. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the Davis-Besse Main Steam Nozzle-to-Shell Weld (Case ID SGB-P3A from Reference [9.1])

Time (yr)

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 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 inspection scenario of PSI+10+40+70, a nozzle 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 used, which represents the highest stress multiplier that can be applied without exceeding the acceptance criterion of 1.0x10-6. The results of the evaluation are summarized in Table 6 up to 80 years of plant operation for the PSI/ISI inspection scenario of (PSI+10+40+70).

L-24-233 Page 12 of 26 Table 6. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the Remaining Davis-Besse SG Welds (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 the Davis-Besse SG indicates that with conservative inputs of the critical parameters, the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-6 failures per year. The stress multipliers applied to Tables 4 through 6 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 Davis-Besse. The evaluation incorporates conservative assumptions with regard to the PSI/ISI scenarios. Furthermore, the evaluation was performed for 30 years, which is longer than the deferral being sought by Vistra OpCo in this request for alternative.

An evaluation was performed to show acceptability of the low KIC values at the beginning and ending of the heatup/cooldown transient for the FW and MS nozzles to address Item No. 2.e.iii during the NRC audit of PROMISE [9.13]. The evaluation was performed using an RTNDT value of 60°F, the maximum allowed by BTP 5-3 [9.14]. The RTNDT value of 60°F bounds the 0°F value in Attachment 2 for the SG materials at Davis-Besse. The evaluation showed acceptable results for the limiting Case IDs from the Reference [9.1]

EPRI report. This was found acceptable by the NRC [9.15]. A similar evaluation was performed for the remainder of the SG welds in Reference [9.16] using the limiting Case ID from the Reference [9.2] EPRI report to address NRC RAI-6 in Reference [9.17]. In this evaluation, the limiting RTNDT value of 60°F was used and acceptable results were also obtained.

L-24-233 Page 13 of 26 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 Davis-Besse, Vistra OpCo does not use Appendix VIII procedures for all the examination categories included in the request for alternative. Based on the observations made by the NRC in Section 3.8.8.2, page 21 of the Vogtle SE [9.7], 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 Davis-Besse by demonstrating that it takes significantly more than 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.

Pressurizer Plant-specific DFM and PFM evaluations were performed for the ONS1/2/3 B&W design pressurizers using the results of the stress analyses in Attachment 5. The DFM and PFM evaluations are presented in Attachment 8 of Reference [9.12]. For convenience, of Reference [9.12] is attached as Attachment 6 of this request for alternative. The information in Table 3 (above) shows that the key plant-specific parameters for the Davis-Besse B&W pressurizer are consistent with those of the ONS1/2/3 pressurizers used in the evaluation in Attachments 5 and 6. Therefore, the DFM and PFM evaluations in Attachment 6 demonstrate that all plant-specific requirements are met for Davis-Besse.

The technical approach used in the DFM evaluation for the B&W pressurizer is consistent with Section 8.2 in the Reference [9.3] EPRI report. The design inputs used in the DFM evaluation are summarized in Table 1 of Attachment 6. An initial flaw size of 5.2 percent of the wall thickness was assumed, equivalent to the most conservative ASME Code,Section XI acceptance standard for these components. The ASME Code,Section XI, Appendix A, Paragraph A-4300 fatigue crack growth (FCG) law was used in the evaluation using the through-wall stress distributions from the stress analyses in. In addition, the weld residual stress from Figure 8-1 in the Reference [9.3]

EPRI Report and the 30 ksi clad residual stress discussed in Section 8.2.2.4 of the Reference [9.3] EPRI report were considered in the evaluation. The fracture mechanics models identified in Section 8.2.2.4 of the Reference [9.3] EPRI report were used to determine the length of time for the postulated initial flaw to grow to a depth of 80 percent of the wall thickness (assumed to equate to leakage in this evaluation) or the depth at which the allowable toughness (upper shelf value of KIC equal to 106 ksiinch reduced by L-24-233 Page 14 of 26 a structural factor of 2.0 for primary stresses and 1.0 for secondary stresses) was reached, whichever was less.

The results of the DFM evaluation for the B&W pressurizer configuration are summarized in Table 4 of Attachment 6, which shows that for the DFM evaluation the period required for hypothetical postulated flaws to leak are very long (in excess of 200 years). This indicates that the B&W pressurizer welds are very flaw tolerant. Because the DFM evaluation considered hypothetical postulated flaws, structural factors of 2.0 on primary loads and 1.0 on secondary loads, consistent with ASME Code,Section XI, Appendix G, were applied.

The PFM evaluations in Attachment 6 were performed consistently with the approach described in Section 8.3 of the Reference [9.3] EPRI report using PROMISE, Version 2.0.

The design inputs used for the PFM evaluation are shown in Table 5 of Attachment 6. For the Davis-Besse pressurizer, PSI examinations have been performed followed by ISI examinations over four completed 10-year intervals (Davis-Besse is currently in the first period of the fifth inspection interval). The PSI/ISI scenario is therefore PSI plus four 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+30+40+70).

This PSI/ISI scenario is consistent with the scenario evaluated in Attachment 5.

Therefore, the PFM results of Attachment 6 are directly applicable to the Davis-Besse pressurizer.

In Attachment 6, stress and fracture toughness were identified as the key variables in the PFM evaluation in Reference [9.3]. As such, three sensitivity studies were performed as part of the PFM as follows:

1. The fracture toughness was decreased to determine the minimum fracture toughness that will meet the acceptance criteria of 1.0x10-6.

2. The stresses were increased to determine the maximum stress multiplier that will meet the acceptance criteria of 1.0x10-6.

3. A sensitivity study of the combined effects of the fracture toughness and stress.

Inspection history for the Davis-Besse pressurizer components is provided in Table 2-3 of. For Item Nos. B2.11 and B2.12, inspection coverage is greater than 90 percent (essentially 100 percent) for all welds. However, for Items B3.110, some welds have limited coverage. The minimum coverage at Davis-Besse is 52.48 percent. A sensitivity study was performed for a coverage of 25.2 percent in Attachment 6, which is less than the minimum coverage of 52.48 percent at Davis-Besse. Evaluations were performed using this limiting coverage to determine the probabilities of rupture and leakage for the plant-specific inspection scenarios of (PSI+10+10+30+40+70) using the same input parameters as in Table 6 of Attachment 6. For comparison, evaluations were L-24-233 Page 15 of 26 also performed for the current ASME Code,Section XI mandated 10-year inspection interval of (PSI+10+20+30+40+50+60+70).

The results of the PFM evaluation are presented in Table 6 of Attachment 6. As shown in this table, the probabilities of rupture and leakage are all below the acceptance criteria of 1.0x10-6 after 80 years of plant operation by three orders of magnitude.

The results of the sensitivity studies are presented in Tables 7 through 9 of Attachment 6.

From Table 7 of Attachment 6, the fracture toughness can be as low as 72 ksiin before the acceptance criterion of 1.0x10-6 is reached after 80 years of operation. From Table 8 of Attachment 6, a stress multiplier of 1.4 can be applied to all the stresses considered in the evaluation before the acceptance criterion is reached. Table 9 of Attachment 6 shows that by applying a stress multiplier of 1.1 and reducing the fracture toughness to 80 ksiin, the probabilities of rupture and leakage are all below the acceptance criterion of 1.0x10-6 after 80 years of plant operation. These sensitivity studies demonstrate the additional margins that are inherent in the PFM evaluation.

The results of the sensitivity study on coverage are presented in Table 10 of Attachment

6. As shown in this table, considering a coverage of 25.2 percent (which is less than the minimum coverage of 52.48 percent at Davis-Besse) for Item No. B3.110 and the Davis-Besse PSI/ISI scenario, the probabilities of rupture and leakage are below the acceptance criteria of 1.0x10-6 after 80 years of operation by three orders of magnitude. Furthermore, when the probabilities of rupture and leakage for the alternative inspection schedule are compared to the present ASME Code,Section XI inspection schedule, there is no difference. This indicates that there is no change in risk from the current ASME Code,Section XI schedule to that of the alternative inspection schedule.

The DFM evaluation in Table 8-4 of Reference [9.3] provides verification of the above PFM results for Davis-Besse 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], Section 8.3.4.1 of Reference [9.2],

and Section 8.3.4.1 of Reference [9.3], preservice examination (PSI) refers to the collective examinations required by ASME Code,Section III during fabrication and any ASME Code,Section XI examinations performed prior to service. The Section III fabrication examinations required for these components were robust and any Section XI preservice examinations further contributed to thorough initial examinations.

Inspection history for the Davis-Besse SG (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in. As shown in the attachment, all examinations obtained an inspection coverage of greater than 90 percent (essentially 100 percent), and no flaws that L-24-233 Page 16 of 26 exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

Inspection history for the Davis-Besse pressurizer (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in. As shown in the attachment, some of the welds have limited exam coverage, with the minimum coverage being 52.48 percent. As shown in Attachment 3, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

Industry Survey The inspection history for these components as obtained from an industry survey is presented in Attachment 4. The results of the survey indicate that these components are very flaw tolerant.

Performance Monitoring The proposed examination schedule for Davis-Besse SG and pressurizer welds and components is presented in the tables on page 5 of this request for alternative. As seen in these tables, all SG and pressurizer welds and components (100 percent of the required ASME section examinations) at Davis-Besse will be inspected approximately 20 years from the date of the last inspection. This is considerably more than the 25 percent sampling during the deferral period required by the NRC for adequate performance monitoring using the binomial distribution model in ML23033A667 [9.39] and ML23114A034 [9.40].

If during the examination schedule described above indications are detected that exceed the applicable ASME Code,Section XI acceptance standards of IWB-3500 or IWC-3500, the indications will be addressed as required by ASME Code,Section XI. The additional examination and successive inspection requirements of ASME Code,Section XI, will be applied. The number of additional exams shall be the number required by ASME Code,Section XI, IWB-2430/IWC-2430.

In addition to ASME Code,Section XI, Davis-Besse utilizes the Corrective Action Program to review and evaluate industry Operating Experience (OE) to determine the appropriate actions required based upon the specific OE. If the OE indicates that a new or novel degradation mechanism may exist in SG and pressurizer welds or components, appropriate examinations will be performed to ensure that no such mechanism is occurring at Davis-Besse.

Conclusion It is concluded that the pressure-retaining welds and full penetration welded nozzles of the Davis-Besse SGs and pressurizer are very flaw tolerant. PFM and DFM evaluations performed as part of technical basis reports [9.1], [9.2] and [9.3], as supplemented by plant-specific evaluations performed as part of this request for alternative, demonstrate L-24-233 Page 17 of 26 that using conservative PSI/ISI inspection scenarios for all plants, the NRC safety goal of 1.0x10-6 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to Davis-Besse SGs and pressurizer is demonstrated in Attachments 2 and 3, respectively. 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 2 and 3 show the examination history for the Davis-Besse SG and pressurizer welds and components examined in the most recent 10-year inspection interval (SG) and the two most recent 10-year inspection intervals (pressurizer).

In addition to the required PSI examinations for these SG and pressurizer welds and components, Davis-Besse has performed multiple ISI examinations through the current 10-year inspection interval.

No flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations, as shown in Attachments 2 and 3.

Some examinations listed in Attachment 3 have limited examination coverage (less than 90 percent). With regard to the Davis-Besse B&W pressurizer design, evaluations were performed in Attachment 8 of Reference [9.12] to show that the probabilities of rupture and leakage for the ISI scenarios for a B&W pressurizer (ONS1/2/3) are similar to those corresponding to performing the regular Section XI inspections every 10 years. This is consistent with Section 8.3.5 of Reference [9.3], which discusses limited coverage and determined that the conclusions of the reports are applicable to components with limited coverage.

In addition, it is important to note all other inspection activities, including the system leakage test (Examination Categories B-P and C-H) will continue to be performed in accordance with the ASME Code,Section XI requirements, providing further assurance of safety.

A performance monitoring plan has been presented to address the unlikely occurrence of a novel degradation mechanism.

Finally, as discussed in Reference [9.18], for situations where no active degradation mechanism is present, Vistra OpCo 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, Vistra OpCo requests the NRC to grant this proposed alternative in accordance with 10 CFR 50.55a(z)(1).

L-24-233 Page 18 of 26

6.

Duration of Proposed Alternative The proposed alternative is requested for the remainder of the fifth 10-year inspection interval and through the sixth 10-year inspection interval for Davis-Besse. The sixth 10-year inspection interval is currently scheduled to end on September 20, 2042, recognizing that the current 60-year operating license expires April 22, 2037.

The proposed alternative is to increase the inspection interval for the applicable SG and pressurizer Item Nos. from the current ASME Code,Section XI 10-year requirement, thereby deferring examinations for two 10-year ISI intervals from the last examination performed for each Item No. The subject welds will be reexamined prior to the end of the current 60-year operating license for Davis-Besse.

7.

Precedents The following previous submittal has been made by SNC to provide relief from the ASME Code,Section XI Examination Category C-B (Item Nos. C2.21 and C2.22) surface and volumetric examinations based on the Reference [9.1] technical basis report:

Letter from C. A. Gayheart (SNC) to NRC, Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Alternative VEGP-ISI-ALT-04-04 Version 2.0, dated September 9, 2020 (ADAMS Accession No. ML20253A311), Reference [9.21].

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

Letter from M. T. Markley (NRC) to C. A. Gayheart (SNC), Vogtle Electric Generating Plant, Units 1 and 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code (EPID L-2020-LLR-0109), dated January 11, 2021 (ADAMS Accession No. ML20352A155), Reference [9.7].

The following previous submittal has been made by Dominion Energy Nuclear Connecticut, Inc. (Dominion) 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 M. D. Sartain (Dominion) to NRC, Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles, dated July 15, 2020 (ADAMS Accession No. ML20198M682), Reference [9.22].

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

L-24-233 Page 19 of 26 Letter from J. G. Danna (NRC) to D. G. Stoddard (Dominion), Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No.

RR-05-06 (EPID L-2020-LLR-0097), dated July 16, 2021 (ADAMS Accession No. ML21167A355), Reference [9.8].

The following previous submittal has been made by PSEG Nuclear LLC (PSEG) to provide relief from the ASME Code,Section XI, Examination Category B-B (Item Nos.

B2.11 and B2.12) and Category B-D (Item No. B3.110) surface and volumetric examinations based on the Reference [9.3] technical basis report:

Letter from P. R. Duke, Jr. (PSEG) to NRC, Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12, dated August 5, 2020 (ADAMS Accession No. ML20218A587), Reference [9.23].

The NRC issued a safety evaluation of the PSEG request for alternative on June 10, 2021.

Letter from J. G. Danna (NRC) to E. Carr (PSEG), Salem Generating Station Unit Nos. 1 and 2 - Authorization and Safety Evaluation for Alternative Request No.

SC-I4R-200 (EPID L-2020-LLR-0103), dated June 10, 2020 (ADAMS Accession No. ML21145A189), Reference [9.9].

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 Relief Requests Associated with the First and Second 10-Year Interval of the Inservice Inspection (ISI) Plan, Millstone Nuclear Power Station, Unit No. 3 (TAC No. MA5446), dated July 24, 2000 (ADAMS Accession No. ML003730922).

Letter from R. L. Emch (NRC) to J. B. Beasley, Jr. (Southern Nuclear Operating Company, Inc.), 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), dated June 20, 2001 (ADAMS Accession No. ML011640178).

Letter from T. H. Boyce (NRC) to C. L. Burton (Carolina Power & Light Company),

Shearon Harris Nuclear Power Plant, Unit 1 - Requests for Relief 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, and 2R2-011 for the Second 10-Year Interval Inservice Inspection Program Plan (TAC Nos. ME0609, ME0610, ME0611, ME0612, ME0613, ME0614, and ME0615), dated January 7, 2010 (ADAMS Accession No. ML093561419).

Letter from M. Khanna (NRC) to D. A. Heacock (Dominion), Millstone Power Station, 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 L-24-233 Page 20 of 26 Through ME6006), dated March 12, 2012 (ADAMS Accession No. ML120541062).

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

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

Letter from M. G. Kowal (NRC) to M. A. Balduzi (Entergy Nuclear Operations, Inc.),

Indian Point Nuclear Generating Unit No. 2 - Relief Request No. RR-01 (TAC No.

MD4695), dated September 5, 2007 (ADAMS Accession No. ML072130487).

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

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

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

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

Catawba Nuclear Station, Unit 1 - Request for Relief 05-CN-004, Limited Weld Examinations During End-of-Cycle 15 Refueling Outage (TAC Nos. MC8337, MC9171, MC9172, MC9173, MC9174, MC9175, MC9176, MC9177, MC9178, and MC9179), dated September 25, 2006 (ADAMS Accession No. ML062390020).

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

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

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

Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 14-CN-001, American Society of Mechanical Engineers (ASME)Section XI Volumetric L-24-233 Page 21 of 26 Examination Requirements (TAC Nos. MF3527 AND MF3528), dated October 30, 2014 (ADAMS Accession No. ML14295A532).

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

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

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

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

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

Based on studies presented in Reference [9.24], the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference [9.25].

Based on work performed in BWRVIP-108 [9.6] and BWRVIP-241 [9.28], the NRC approved the reduction of BWR vessel FW nozzle-to-shell weld examinations (Item No. B3.90 for BWRs from 100 percent to a 25 percent sample of each nozzle type every 10 years) in References [9.27] and [9.29]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702

[9.30], which has been conditionally approved by the NRC in Revision 19 of Regulatory Guide 1.147 [9.31].

L-24-233 Page 22 of 26

8.

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 ID Inside diameter ISI Inservice Inspection MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system OD Outside diameter PDI Probability of detection PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor PZR Pressurizer SCC Stress corrosion cracking SG Steam Generator WEC Westinghouse Electric Company

9.

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 (ADAMS Accession No. ML19347B107).

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, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906 (ADAMS Accession No. ML20225A141).

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

L-24-233 Page 23 of 26 9.4 N. Palm (EPRI), BWR Vessel & Internals Project (BWRVIP) Memo No. 2019-016, White Paper on Suggested Content for PFM Submittals to the NRC, February 27, 2019 (ADAMS Accession No. ML19241A545).

9.5 NRC Regulatory Guide 1.245, Revision 0, Preparing Probabilistic Fracture Mechanics Submittals, January 2022.

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

9.7 Letter from M. T. Markley (NRC) to C. A. Gayheart (SNC), Vogtle Electric Generating Plant, Units 1 & 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code (EPID L-2020-LLR-0109), dated January 11, 2021 (ADAMS Accession No. ML20352A155).

9.8 Letter from J. G. Danna (NRC) to D. G. Stoddard (Dominion), Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No.

RR-05-06 (EPID L-2020-LLR-0097), dated July 16, 2021 (ADAMS Accession No. ML21167A355).

9.9 Letter from J. G. Danna (NRC) to E. Carr (PSEG Nuclear), Salem Generating Station Unit Nos. 1 and 2 - Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200 (EPID L-2020-LLR-0103), dated June 10, 2020 (ADAMS Accession No. ML21145A189).

9.10 Email Letter from R. Guzman (NRC) to S. Sinha (Dominion), 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), dated February 3, 2021 (ADAMS Accession No. ML21034A576).

9.11 Letter from G. T. Bischof (Dominion), 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, dated March 19, 2021 (ADAMS Accession No. ML21081A136).

9.12 Letter from K. Ellis (Duke Energy) to the NRC on February 17, 2023,

Subject:

Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) (ADAMS Accession No. ML23048A148).

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

L-24-233 Page 24 of 26 9.14 NUREG-0800 - Chapter 5, Branch Technical Position (BTP) 5-3, Revision 2, Fracture Toughness Requirements.

9.15 Letter from J. G. Lamb (NRC) to C. A. Gayheart (SNC), Vogtle Electric Generating Plant, Units 1 and 2 - Audit Report for the PROMISE Version 1.0 Probabilistic Fracture Mechanics Software Code Used in Relief Request VEGP-ISI-ALT-04-04 (EPID L-2020-LLR-0109), dated December 10, 2020 (ADAMS Accession No. ML20258A002).

9.16 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, dated June 17, 2022 (ADAMS Accession No. ML22168A005).

9.17 Email Letter from J. Wiebe (NRC) to T. Loomis (Constellation Energy Generation, LLC), Draft RAIs for Requests for Alternatives I4R-17, I4R-23, ISI-05-018, I6R-10 (EPID Nos.: L-2021-LLR-091, L-2021-LLR-092, L-2021-LLR-093, L-2021-LLR-094), dated May 6, 2022 (ADAMS Accession No. ML22129A013).

9.18 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.

9.19 Not used.

9.20 Not used.

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

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

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

9.24 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.

L-24-233 Page 25 of 26 9.25 NRC, Revised Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, Pressurized Water Reactor Owners Group, Project No. 694, July 26, 2011 (ADAMS Accession No. ML111600303).

9.26 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.

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

9.28 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.

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

9.30 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.

9.31 NRC Regulatory Guide 1.147, Revision 18, Inservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1, dated March 2017.

9.32 Davis-Besse Drawing No. M-506-00188-2 (B&W Drawing No. 205SE001 Rev 06),

Davis-Besse ROTSG General Arrangement.

9.33 Davis-Besse Drawing No. M-506-00450-1 (B&W Drawing No. 205SE138 Rev 03),

Steam Outlet Nozzle.

9.34 Davis-Besse Drawing No. ISI-SK-020, Pressurizer Outline, Containment Building, Revision 2.

9.35 Davis-Besse Document, Updated Safety Analysis Report, Davis-Besse Nuclear Power Station No. 1, Docket No: 50-346, License No: NPF-3, Revision 34b.

9.36 Davis-Besse Drawing No. M-507-26-3 (B&W Drawing No. 154595E), Lower Head, Assembly & Details, Revision T2.

9.37 Davis-Besse Drawing No. M-507-32-1 (B&W Drawing No. 154599), Heater Belt Details, Revision 2.

L-24-233 Page 26 of 26 9.38 Davis-Besse Drawing No. M-507-74-3 (B&W Drawing No. 154600E), Surge Nozzle Details, Revision T2.

9.39 NRC Presentation, Performance Monitoring, January 30, 2023 (ADAMS Accession No. ML23033A667).

9.40 NRC Presentation, Probabilistic Fracture Mechanics and Performance Monitoring, NRC Public Meeting, April 27, 2023 (ADAMS Accession No. ML23144A034).

L-24-233 Plant-Specific Applicability Davis-Besse Steam Generator L-24-233 Page 1 of 11 Section 9 of References [1-1] and [1-2] provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Davis-Besse 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 Davis-Besse.

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

Category Requirement from Reference [1-1]

Applicability to Davis-Besse 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 portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance.

For the replacement SGs that were installed in 2014 and are currently in service, Davis-Besse 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 blackout, resulting in thermal shock of any portion of the vessel.

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

The Davis-Besse SG vessel heads and tubesheet are fabricated of SA-508, Grade 3, Class 2 material (Reference [1-3]). The RTNDT values for the Davis-Besse SG vessel heads and tubesheet materials are 0°F or less (Reference [1-4]) (so the RTNDT of 60°F used in the EPRI report is bounding).

This 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 1-1 and Figure 1-2 of Reference [1-1].

The Davis-Besse tubesheet-to-head weld configuration is shown in Figure 1-2 below 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 main section of this request for alternative, the Davis-Besse SG vessel dimensions are as follows:

SG Lower Head OD = 131.81" L-24-233 Page 2 of 11 Category Requirement from Reference [1-1]

Applicability to Davis-Besse SG Upper Shell OD = 144.125 The upper shell dimension is within 10% of that specified in Table 9-2 in Section 9.4.3 of Reference [1-1] for B&W plants (Reference [1-5]). The lower head is within 11.4% of that specified in Table 9-2. The Table 9-2 lower head diameter was assumed the same as the upper shell, but did not account for the reduction in diameter of the head. Upon comparison with Figure 4-3 of Reference [1-1], it can be seen that the head dimension is consistent with that of the B&W design evaluated and is therefore deemed to be within acceptable geometrical tolerances.

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 Davis-Besse 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].

Item No. C1.30 (SG Secondary Side Shell Welds)

Category Requirement from Reference [1-1]

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

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

For the replacement SGs that were installed in 2014 and are currently in service, Davis-Besse 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 blackout, resulting in thermal shock of any portion of the vessel.

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

The Davis-Besse SG vessel shell and tubesheet are fabricated of SA-508, Grade 3, Class 2 material (Reference

[1-3]). The RTNDT values for the Davis-Besse SG vessel shell and tubesheet L-24-233 Page 3 of 11 Category Requirement from Reference [1-1]

Applicability to Davis-Besse materials are 0°F or less (Reference

[1-4]) (so the RTNDT of 60°F used in the EPRI report is bounding).

This 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 Davis-Besse weld configuration is shown in Figure 1-3 and conforms to Figure 1-8 of Reference [1-1]. (Note:

Vistra OpCo is not requesting relief for Item No. C1.10 or C1.20 components shown in Figure 1-7 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 9.4.4 of Reference [1-1].

Per Table 1 in the main section of this request for alternative, the Davis-Besse SG vessel dimensions are as follows:

SG Lower Head OD = 131.81 SG Upper Shell OD = 144.125 The upper shell dimension is within 10% of that specified in Table 9-2 in Section 9.4.3 of Reference [1-1] for B&W plants (Reference [1-5]). The lower head is 11.4% of that specified in Table 9-2. The Table 9-2 lower head diameter was assumed the same as the upper shell, but did not account for the reduction in diameter of the head.

Upon comparison with Figure 4-3 of Reference [1-1], it can be seen that the head dimension is consistent with that of the B&W design evaluated and is therefore deemed to be within acceptable geometrical tolerances.

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 Davis-Besse 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].

L-24-233 Page 4 of 11 Item Nos. C2.21 and C2.22 (SG MS Nozzle-to-Shell Welds and Inside Radius Sections)

Category Requirement from Reference [1-2]

Applicability to Davis-Besse 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 Davis-Besse MS nozzle-to-shell weld is shown in Figure 1-4 below and is representative of the configuration shown in Figure 1-2 of Reference [1-2].

The materials of the SG shell 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 Davis-Besse SG shell and MS nozzles are fabricated of SA-508, Grade 3, Class 2 material (Reference

[1-3]). The RTNDT value for the material of Davis-Besse SG nozzle-to-shell welds is 0°F or less (Reference [1-5])

(so the RTNDT of 60°F used in the EPRI report is bounding).

This material is a low alloy ferritic steel that 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 Davis-Besse 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 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 Davis-Besse (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 Davis-Besse MS nozzle is 24 Sch. 60 (Reference [1-5]). (Note: There is also a 26x24 reducer between the nozzle and attached piping.)

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.

Davis-Besse is a B&W design, with the main steam nozzle exiting the side of the SG.

L-24-233 Page 5 of 11 Category Requirement from Reference [1-2]

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

The Davis-Besse MS nozzle configuration (shown in Figure 2-4) does not protrude significantly into the SG as shown in Figure 4-7 of Reference [1-2] and does not have a unique weld configuration as shown in Figure 4-6 of Reference [1-2]

(Reference [1-5]).

Table 1-2. Davis-Besse Data for Thermal Transients for Stress Analysis of the 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]

Davis-Besse 60-Year Projection Heatup / Cooldown 300 114 / 114(2)

Plant Loading / Unloading 5,000 1,800 / 1,800(3)

Reactor Trip 360 187(4)

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.

Transient Numbers 1A and 1B in Attachment 1 (page 1 of 6) of Reference [1-7] (Note:

Values provided are best estimate).

3.

Transient Numbers 3 and 4 in Attachment 1 (Page 1 of 6) of Reference [1-7].

4.

Transients Number 8A, 8B, 8C and 8D in Attachment 1 (Page 1 of 6) of Reference [1-7]

(Note: Transient Number 8E is an Emergency transient).

L-24-233 Page 6 of 11 Table 1-3. Davis-Besse Data for Thermal Transients for Stress Analysis of the 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]

Davis-Besse 60-Year Projection Heatup / Cooldown 300 114 / 114(2)

Plant Loading / Unloading 5,000 1,800 / 1,800(3)

Reactor Trip 360 187(4)

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.

Transient Numbers 1A and 1B in Attachment 1 (page 1 of 6) of Reference [1-7] (Note:

Values provided are best estimate).

3.

Transient Numbers 3 and 4 in Attachment 1 (Page 1 of 6) of Reference [1-7].

4.

Transients Number 8A, 8B, 8C and 8D in Attachment 1 (Page 1 of 6) of Reference [1-7]

(Note: Transient Number 8E is an Emergency transient).

Table 1-4. Davis-Besse Data for Thermal Transients Applicable to SG MS 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]

Davis-Besse 60-Year Projection Heatup / Cooldown 300 114 / 114(1)

Plant Loading 5,000 1,800(2)

Plant Unloading 5,000 1,800(3)

Loss of Load 360 187(4)

Loss of Power 60 6(5)

Notes:

1.

Transient Numbers 1A and 1B in Attachment 1 (page 1 of 6) of Reference [1-7]

(Note: Values provided are best estimate).

2.

Transient Number 3 in Attachment 1 (Page 1 of 6) of Reference [1-7].

3.

Transient Number 4 in Attachment 1 (Page 1 of 6) of Reference [1-7].

4.

Transients Number 8A, 8B, 8C and 8D in Attachment 1 (Page 1 of 6) of Reference [1-7] (Note: Transient Number 8E is an Emergency transient).

5.

Transient Number 15 in Attachment 1 (page 2 of 6) of Reference [1-7].

L-24-233 Page 7 of 11 Table 1-5. Davis-Besse SG Inspection History ASME Category Item No.

Exam Date Interval / Period (Outage)

Component ID Exam Results(1)

Coverage Relief Request B-B B2.40 8/21/2013 4th Interval/ 1st Period (PSI for 18R)

RCSG11W23 NRI 97.0%

N/A B-B B2.40 3/20/2022 4th Interval / 3rd Period (22R)

RCSG11W23 NRI 97.0%

N/A B-B B2.40 8/21/2013 4th Interval / 1st Period (PSI for 18R)

RCSG12W23 NRI 97.0%

N/A B-B B2.40 8/21/2013 4th Interval / 1st Period (PSI for 18R)

RCSG11W22 NRI 99.0%

N/A B-B B2.40 8/21/2013 4th Interval / 1st Period (PSI for 18R)

RCSG12W22 NRI 99.0%

N/A B-B B2.40 3/19/2018 4th Interval / 2nd Period (20R)

RCSG12W22 NRI 96.6%

N/A C-A C1.30 8/22/2013 4th Interval / 1st Period (PSI for 18R)

SPSG11W65 NRI 100.0%

N/A C-A C1.30 8/21/2013 4th Interval / 1st Period (PSI for 18R)

SPSG11W69 NRI 99.8%

N/A C-A C1.30 3/20/2022 4th Interval / 3rd Period (22R)

SPSG11W69 NRI 97.0%

N/A C-A C1.30 8/7/2013 4th Interval / 1st Period (PSI for 18R)

SPSG12W65 NRI 99.0%

N/A C-A C1.30 3/19/2018 4th Interval / 2nd Period (20R)

SPSG12W65 NRI 100.0%

N/A C-A C1.30 8/7/2013 4th Interval / 1st Period (PSI for 18R)

SPSG12W69 NRI 96.0%

N/A C-B C2.21 8/21/2013 4th Interval / 1st Period (PSI for 18R)

SPSG11W127 X/Y NRI 100.0%

N/A C-B C2.21 8/21/2013 4th Interval / 1st Period (PSI for 18R)

SPSG11W128 W/X NRI 100.0%

N/A C-B C2.21 3/13/2020 4th Interval / 3rd Period (21R)

SPSG11W128 W/X NRI 100.0%

N/A C-B C2.21 8/21/2013 4th Interval / 1st Period (PSI for 18R)

SPSG12W127 X/Y NRI 100.0%

N/A C-B C2.21 3/17/2018 4th Interval / 2nd Period (20R)

SPSG12W127 X/Y NRI 100.0%

N/A C-B C2.21 8/21/2013 4th Interval / 1st Period (PSI for 18R)

SPSG12W128 W/X NRI 100.0%

N/A C-B C2.22 8/20/2013 4th Interval / 1st Period (PSI for 18R)

SPSG11W127 X/YIR NRI 100.0%(2)

N/A C-B C2.22 8/20/2013 4th Interval / 1st Period (PSI for 18R)

SPSG11W128 W/X-IR NRI 100.0%(2)

N/A C-B C2.22 3/7/2020 4th Interval / 3rd Period (21R)

SPSG11W128 W/X-IR NRI 100.0%(2)

N/A C-B C2.22 7/31/2013 4th Interval / 1st Period (PSI for 18R)

SPSG12W127 X/YIR NRI 100.0%(2)

N/A C-B C2.22 3/19/2018 4th Interval / 2nd Period (20R)

SPSG12W127 X/YIR NRI 100.0%(2)

N/A C-B C2.22 7/31/2013 4th Interval / 1st Period (PSI for 18R)

SPSG12W128 W/X-IR NRI 100.0%(2)

N/A Notes:

1.

NRI = no reportable indications.

2.

100% of area defined in EPRI Report IR-2011-426, Davis-Besse ROTSG Nozzle Examination.

L-24-233 Page 8 of 11 Figure 1-1. Davis-Besse Steam Generator Layout [1-3, 1-5]

L-24-233 Page 9 of 11 Figure 1-2. Davis-Besse Item No. B2.40 Weld Configuration [1-3]

L-24-233 Page 10 of 11 Figure 1-3. Davis-Besse Item No. C1.30 Weld Configuration [1-3]

L-24-233 Page 11 of 11 Figure 1-4. Davis-Besse Main Steam Nozzle Configuration [1-6]

References 1-1.

Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906 (ADAMS Accession No. ML20225A141).

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 (ADAMS Accession No. ML19347B107).

1-3.

Davis-Besse Drawing M-506-00190,Section XI Pre-Service NDE Examination, Revision 1 (Vendor Drawing No. 205SE007, Revision 2).

1-4.

Certified Design Specification TS-3985, Revision 3, for Replacement Steam Generators for FirstEnergy Corporation Davis-Besse Nuclear Power Station, March 2014 (Section 4.12.15).

1-5.

Davis-Besse Drawing M-506-00323, Steam Outlet Nozzle Ordering, Revision 1 (Vendor Drawing No. 205SC136, Revision 1).

1-6.

Davis-Besse Drawing M-506-00188, Davis Besse ROTSG General Arrangement, Revision 2 (Vendor Drawing No. 205SE001, Revision 6).

1-7.

Davis-Besse Procedure EN-DP-00355, Determination of Allowable Operating Transient Cycles, Revision 10.

L-24-233 Plant-Specific Applicability Davis-Besse Pressurizer L-24-233 Page 1 of 4 Section 9 of Reference [2-1] provides requirements that must be demonstrated to apply the representative stress and flaw tolerance analyses to a specific plant. However, the B&W pressurizer configuration at Davis-Besse is considerably different in terms of both geometry and materials from the configuration evaluated in Reference [2-1]. Therefore, in lieu of comparison to the requirements of Section 9 of Reference [2-1], plant-specific stress analyses and fracture mechanics (DFM and PFM) evaluations were performed for a representative B&W design pressurizer (at ONS1/2/3) in Attachments 7 and 8 of Reference [2-2], which are provided as Attachments 5 and 6 of this request for alternative. Additionally, Tables 2-1 through 2-2 (below) demonstrate that the general transients and insurge/outsurge transients at Davis-Besse, respectively, are comparable to those used in the Attachment 5 and 6 evaluations. Table 2-3 provides the inspection history of the Davis-Besse pressurizer components.

Table 2-1. Comparison of Davis-Besse General Transients to Those Used in Attachments 5 and 6 Transient Number of Cycles for 60 Years used in Attachments 5 and 6 Davis-Besse 60-Year Projection Heatup /

Cooldown 300 114 / 114(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 360 187(2)

Notes:

1.

Transient Numbers 1A and 1B in Attachment 1 (page 1 of 6) of Reference [2-3] (Note: Values provided are best estimate).

2.

Transients Number 8A, 8B, 8C and 8D in Attachment 1 (Page 1 of 6) of Reference [2-3] (Note: Transient Number 8E is an Emergency transient).

L-24-233 Page 2 of 4 Table 2-2. Comparison of Davis-Besse Insurge/Outsurge Transients to Those Used in Attachments 5 and 6 T (oF)(1) 60-Year No. of Cycles used in Attachments 5 and 6 Davis-Besse 60-Year Projection [2-4]

400 100 0

350 1,500 45 300 1,500 34 250 1,500 45 200 6,500 1704 Notes:

1.

T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

L-24-233 Page 3 of 4 Table 2-3. Davis-Besse Pressurizer Inspection History ASME Cat.

Item No.

Exam Date Interval / Period (Outage)

Component ID Exam Results(1)

Coverage Relief Request B-D B3.110 2/26/2002 3rd Interval / 1st Period (13R)

RC-PZR-WP W/X NRI 65.65%

N/A B-D B3.110 2/26/2002 3rd Interval / 1st Period (13R)

RC-PZR-WP-34 NRI 72.14%

N/A B-D B3.110 1/5/2008 3rd Interval / 3rd Period (15R)

RC-PZR-WP Z/W NRI 59.3%(2)

N/A B-B B2.12 1/3/2008 3rd Interval / 3rd Period (15R)

RC-PZR-WP-1 NRI

> 90%

N/A B-B B2.11 1/4/2008 3rd Interval / 3rd Period (15R)

RC-PZR-WP-76 NRI

> 90%

N/A B-D B3.110 10/12/2011 3rd Interval / 3rd Period (17R)

RC-PZR-WP-15 NRI 72.3%(2)

N/A B-D B3.110 10/17/2011 3rd Interval / 3rd Period (17R)

RC-PZR-WP Y/Z NRI 74.1%(2)

N/A B-B B2.12 10/18/2011 3rd Interval / 3rd Period (17R)

RC-PZR-MK 40-6-WP-7-Y-LU NRI

> 90%

N/A B-B B2.11 10/18/2011 3rd Interval / 3rd Period (17R)

RC-PZR-WP-28 NRI

> 90%

N/A B-D B3.110 2/13/2014 4th Interval / 1st Period (18R)

RC-PZR-WP-34 NRI 67.6%

N/A B-D B3.110 2/12/2014 4th Interval / 1st Period (18R)

RC-PZR-WP W/X NRI 52.48%

N/A B-B B2.11 2/11/2014 4th Interval / 1st Period (18R)

RC-PZR-WP-76 NRI

> 90%

N/A B-B B2.12 2/11/2014 4th Interval / 1st Period (18R)

RC-PZR-WP-1 NRI

> 90%

N/A B-D B3.110 3/9/2018 4th Interval / 2nd Period (20R)

RC-PZR-WP Z/W NRI 59.3%

N/A B-D B3.110 3/10/2022 4th Interval / 3rd Period (22R)

RC-PZR-WP Y/Z NRI 74.1%

N/A B-D B3.110 3/16/2022 4th Interval / 3rd Period (22R)

RC-PZR-WP-15 NRI 72.3%

N/A B-B B2.11 3/18/2022 4th Interval / 3rd Period (22R)

RC-PZR-WP-28 NRI 96.3%

N/A B-B B2.12 3/18/2022 4th Interval / 3rd Period (22R)

RC-PZR-MK 40-6-WP-7-Y-LU NRI 91.8%

N/A Notes:

1.

NRI = no reportable indications.

2.

Exam Report only indicates < 90% coverage obtained. Coverage assumed to be identical to that obtained in a later examination of the same component.

L-24-233 Page 4 of 4 References 2-1.

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

2-2.

Letter from K. Ellis (Duke Energy) to the NRC,

Subject:

Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1), dated February 17, 2023 (ADAMS Accession No. ML23048A148).

2-3.

Davis-Besse Procedure EN-DP-00355, Determination of Allowable Operating Transient Cycles, Revision 10.

2-4.

SI Calculation No. 2400544.301, DB Pressurizer Insurge/Outsurge Transients, Revision 0.

L-24-233 Results of Industry Survey L-24-233 Page 1 of 5 Overall Industry Inspection Summary for Code Items B2.31, B2.32, B2.40, B3.130, C1.10, C1.20, and C1.30 The results of an industry survey of past inspections of SG nozzle-to-shell welds, inside radius sections and shell welds are summarized in Reference [3-1]. Table 3-1 provides a summary of the combined survey results for Item Nos. B2.31, B2.32 (see Table 3-1, Note 3), B2.40, B3.130, C1.10, C1.20, and C1.30. The results of the industry survey identified that numerous steam generator (SG) examinations are being performed with no service-induced flaws being detected. Performing these examinations adversely impacts 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 (that is, 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 that required flaw evaluation were identified during these examinations. None of these flaws were found to be service-induced. For Item No. B2.40, examinations at two units at a single plant site identified multiple flaws exceeding the acceptance criteria of ASME Code Section XI; however, these were determined to be subsurface-embedded fabrication flaws and non-service-induced (see Table 3-1, 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 3-1, Note 2). As discussed in References [3-4] and [3-5], these flaws were most likely subsurface weld defects typical of thick vessel welds and not service-induced. A flaw evaluation for IWC-3600 was performed for these flaws, and they were found to be acceptable for continued operation.

L-24-233 Page 2 of 5 Table 3-1. Summary of Survey Results for SG Nozzle-to-Shell, Inside Radius Section, and Shell Weld Components Item No.

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

30 30 0

0 0

B2.32 (Note 3) 0 13 13 0

0 0

B2.40 0

183 183 0

Note 1 Note 1 B3.130 0

135 135 0

0 0

C1.10 140 305 445 0

0 0

C1.20 54 319 373 0

Note 2 Note 2 C1.30 32 163 195 0

0 0

Totals 226 1148 1374 0

Notes 1 and 2 Notes 1 and 2 Notes:

1.

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

2.

A single PWR W-2 Loop unit reported multiple flaws [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.

L-24-233 Page 3 of 5 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. C2.21, C2.22, and C2.32 (see Table 3-2, Note 1). The results identify that SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Section examinations are 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 (that is, B&W, CE, and Westinghouse). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32 (see Table 3-2, 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.

L-24-233 Page 4 of 5 Overall Industry Inspection Summary for Code Items B2.11, B2.12, B2.21, B2.22, and B3.110 The results of an industry survey of past inspections of pressurizer welds are summarized in Reference [3-6]. Table 3-1 provides a summary of the combined survey results for Item Nos.

B2.11, B2.12, B2.21, B2.22, and B3.110. The results identify that pressurizer examination of the items adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international 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 nuclear steam supply system (NSSS) vendors (that is, B&W, CE, and Westinghouse). A total of 1,162 examinations for the components of the affected Item Nos. were conducted on PWR pressurizer components.

A small number of flaws were identified during these examinations, which required flaw evaluation. None of these flaws were found to be service induced. Out of a total of 1,162 examinations identified by the plants that responded to the survey that have been performed on the above item numbers, only four examinations (for Item No. B2.11), at two units of a single plant site, identified flaws exceeding the acceptance criteria of ASME Code,Section XI.

Flaw evaluations were performed to show acceptability of these indications and follow-on examinations showed no change in flaw sizes since the original inspections. No other indications were identified in any in-scope components.

Table 3-3. Summary of Survey Results for Pressurizer Components Item No.

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

B2.12 269 0

B2.21 4

0 B2.22 30 0

B3.110 590 0

Note:

(1) Flaw evaluations were performed to show acceptability of these indications and follow-on examinations showed no change in flaw sizes since the original inspections. None of these flaws were found to be service induced.

L-24-233 Page 5 of 5 References 3-1.

Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906 (ADAMS Accession No. ML20225A141).

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 (ADAMS Accession No. ML19347B107).

3-3.

Letter from F. A. Kearney (Exelon Generation) to NRC, Byron Station Unit 2 90-Day Inservice Inspection Report for Interval 3, Period 3, (B2R17), dated July 29, 2013, Docket No. 50-455 (ADAMS Accession No. ML13217A093).

3-4.

Letter from J. P. Sorensen (Nuclear Management Company, LLC) to NRC, Unit 1 Inservice Inspection Summary Report, Interval 3, Period 3 Refueling Outage Dates 1-19-2001 to 2-25-2001 Cycle 20 / 05-26-99 to 02-25-2001, dated May 29, 2001, Docket Nos. 50-282 and 50-306 (ADAMS Accession No. ML011550346).

3-5.

Letter from J. M. Solymossy (Nuclear Management Company, LLC) 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), dated April 4, 2003, Docket Nos. 50-282 and 50-306 (ADAMS Accession No. ML031040553).

3-6.

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

L-24-233 SI Calculation No. 2100561.302, Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head, Revision 1 (40 pages follow)

CALCULATION PACKAGE File No.: 2100561.302 Project No.: 2100561 Quality Program Type:

Nuclear Commercial PROJECT NAME:

Duke PWR Fleet SG & Pzr Inspection Optimization Relief Requests CONTRACT NO.:

GSA # 03021365 CLIENT:

Duke Energy Corporation PLANT:

Oconee Nuclear Station, Units 1, 2 and 3 CALCULATION TITLE:

Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head Document Revision Affected Pages Revision Description Project Manager Approval Signature & Date Preparer(s) &

Checker(s)

Signatures & Date 0

1 - 37 A A-2 Computer Files Initial Issue Scott Chesworth 3/28/2022 Preparer Richard Bax 3/3/2022 Checker Minghao Qin 3/3/2022 1

1 - 38 A A-2 Computer Files Incorporating Duke Energy Comments Scott Chesworth 8/5/2022 Preparer Richard Bax 8/5/2022 Checker Minghao Qin 8/5/2022

File No.: 2100561.302 Revision: 1 Page 2 of 38 F0306-01R4 Table of Contents 1.0 OBJECTIVE.............................................................................................................. 4 2.0 TECHNICAL APPROACH......................................................................................... 4 2.1 Finite Element Model..................................................................................... 4 2.2 Pressure / Thermal Stress Analyses.............................................................. 4 3.0 DESIGN INPUTS....................................................................................................... 4 4.0 ASSUMPTIONS........................................................................................................ 5 5.0 CALCULATIONS....................................................................................................... 5 5.1 Finite Element Model..................................................................................... 5 5.1.1 Material Properties......................................................................................... 6 5.2 Pressure / Thermal Stress Analysis............................................................... 6 5.2.1 Unit Internal Pressure Loading Analysis......................................................... 6 5.2.2 Thermal Heat Transfer Analyses.................................................................... 6 5.2.3 Thermal Stress Analyses............................................................................... 7 6.0 RESULTS OF ANALYSIS.......................................................................................... 7

7.0 REFERENCES

.......................................................................................................... 7 COMPUTER FILES LISTING...................................................................... A-1 List of Tables Table 1. Material Properties for Carbon Steel, C>0.3% (SA-508, Class 1 or SA-516, Grade 70)............................................................................................................ 9 Table 2. Material Properties for Stainless Steel (SA-240, Type 304).................................. 10 Table 3. Thermal Transients for Pressurizer Surge Nozzle................................................. 11 Table 4. Insurge/Outsurge Transients for Pressurizer Surge Nozzle.................................. 12

File No.: 2100561.302 Revision: 1 Page 3 of 38 F0306-01R4 List of Figures Figure 1. Modeled Dimensions........................................................................................... 13 Figure 2. Weld Locations.................................................................................................... 14 Figure 3. 3-D Finite Element Model.................................................................................... 15 Figure 4. Applied Boundary Conditions and Unit Internal Pressure.................................... 16 Figure 5. Example of Applied Thermal Boundary Conditions for Thermal Transient Analyses............................................................................................................ 17 Figure 6. Example of Applied Thermal Boundary Conditions for Insurge/Outsurge Transient Analyses............................................................................................ 18 Figure 7. Applied Mechanical Boundary Conditions for Thermal Stress Analyses.............. 19 Figure 8. Stress Contours Due to Unit Internal Pressure.................................................... 20 Figure 9. Temperature Contour Heatup/Cooldown Transient (Time = 10,494 seconds)...... 21 Figure 10. Stress Contours of Heatup/Cooldown Transient (Time = 10,494 seconds)........ 22 Figure 11. Temperature Contour Insurge/Outsurge Group 3 Transient (Time = 900 seconds)............................................................................................................ 23 Figure 12. Stress Contours of Insurge/Outsurge Group 3 Transient (Time = 900 seconds)............................................................................................................ 24 Figure 13. Path Locations.................................................................................................. 25 Figure 14. Through-Wall Stress Distribution at Path 1........................................................ 26 Figure 15. Through-Wall Stress Distribution at Path 2........................................................ 27 Figure 16. Through-Wall Stress Distribution at Path 3........................................................ 28 Figure 17. Through-Wall Stress Distribution at Path 4........................................................ 29 Figure 18. Through-Wall Stress Distribution at Path 5........................................................ 30 Figure 19. Through-Wall Stress Distribution at Path 6........................................................ 31 Figure 20. Through-Wall Stress Distribution at Path 7........................................................ 32 Figure 21. Through-Wall Stress Distribution at Path 8........................................................ 33 Figure 22. Through-Wall Stress Distribution at Path 9........................................................ 34 Figure 23. Through-Wall Stress Distribution at Path 10...................................................... 35 Figure 24. Through-Wall Stress Distribution at Path 11...................................................... 36 Figure 25. Through-Wall Stress Distribution at Path 12...................................................... 37 Figure 26. Through-Wall Stress Distribution at Path 13...................................................... 38

File No.: 2100561.302 Revision: 1 Page 4 of 38 F0306-01R4 1.0 OBJECTIVE The objective of this calculation is to develop a finite element model of a typical Babcock and Wilcox (B&W) designed pressurizer water reactor (PWR) pressurizer (PZR) surge nozzle and bottom head and determine stresses due to thermal transients and unit pressure. The through-wall stresses are extracted at the surge nozzle-to-bottom head, bottom head-to-shell weld locations and heater bundle shell thickness region to shell and bottom head welds and are stored in computer files that will be used in a separate fracture mechanics evaluation.

2.0 TECHNICAL APPROACH 2.1 Finite Element Model A finite element model (FEM) is developed using the ANSYS finite element analysis software package [1].

The FEM is a 3-dimensional (3-D) model of a typical B&W pressurizer surge nozzle, bottom head, and lower shell region. The model includes a local portion of the pressurizer lower shell and cladding (which includes the shell thickness increase at the heater bundle region), the pressurizer bottom head and cladding, and the surge nozzle and cladding.

2.2 Pressure / Thermal Stress Analyses Stress analyses are performed for thermal transients and a unit internal pressure. For thermal loads due to thermal transients, thermal analyses are performed to determine the temperature distribution time-histories for each transient. The temperature distributions are then used as input to perform stress analyses for each transient. For internal pressure, an arbitrary unit internal pressure is applied. Due to the linear elastic nature of modeling, the stress results from the unit pressure can be scaled to correspond to actual pressure values as needed. Stress results are saved for use in future evaluations and are listed in Appendix A.

3.0 DESIGN INPUTS A typical B&W PZR surge nozzle, bottom head and lower shell configuration is used as a representative component for the finite element model. The geometry of the surge nozzle and bottom head is derived from Oconee Finite Element Model of Pressurizer Surge Nozzle with Weld Overlay Repair Per Design Dimensions [2]. The dimensions of the lower shell, including the shell thickness increase at the heater bundle region is based on a B&W Pressurizer General Arrangement Drawing [3, 9]. The base metal thickness of the increased thickness region at the heater bundle region is defined as 13.563 inches in Reference [6]. However, Reference [7] indicates that the minimum wall thickness at the heater bundle region is 12.5 inches. For this evaluation, the greater wall thickness of 13.563 inches will be used since it will generate higher thermal stresses which are expected to dominate given the relatively thick component.

The general transients for analysis were previously defined in Table 5-6 Reference [4], while the typical insurge/outsurge transients for Westinghouse (W) and Combustion Engineering (CE) were defined in Table 5-9 of Reference [4]. In Reference [4], the B&W insurge/outsurge transients were derived based on applying scaling factors to the Group 3 insurge/outsurge transients of the Westinghouse/CE insurge/outsurge transients. The same approach is adopted in this calculation and therefore only the Group 3 insurge/outsurge transients for the W/CE pressurizer designs documented in Table 5-9 of Reference [4] will be evaluated.

For the thermal transient stress analyses performed herein, only the transient definitions from Table 5-6 (for the general transients) and Table 5-11 (for the B&W insurge/outsurge transients) of Reference [4]

are required. The number of cycles provided in these tables are not considered herein but will be considered in subsequent deterministic and probabilistic fracture mechanics evaluations.

File No.: 2100561.302 Revision: 1 Page 5 of 38 F0306-01R4 The welds of interest for the lower pressurizer region are identified in ASME Code,Section XI, Table IWB-2500-1 [5], and are as follows:

Item No. B2.11 (Pressurizer Shell Circumferential Weld)

Item No. B2.12 (Pressurizer Shell Longitudinal Weld)

Item No. B3.110 (Pressurizer Nozzle-to-Vessel Welds) 4.0 ASSUMPTIONS A number of assumptions are made during development of the finite element model and the thermal /

pressure stress evaluation, which are listed as follows:

The nozzle-to-lower head, lower head-to-shell and the shell welds are not specifically modeled.

The material properties between the base metals and the weld materials are similar enough that the effect of this assumption will be minimal.

Per References [3, 9] the total circumferential extent of the thickness increase for the heater bundle region is 108°. However, the circumferential extent of the full thickness of this region is not defined and is estimated to be 78°.

Heat transfer coefficients during thermal transients are assumed based on the flow condition for the inside surface of the nozzle and bottom head.

All thermal transients are assumed to start and end at a steady-state uniform temperature.

The stress-free reference temperature for thermal stress calculation is assumed to be an ambient temperature of 70°F, which is used for thermal strain calculations. This assumption is typical for stress analyses in similar components.

All outside surfaces are assumed to be fully insulated and the insulation itself is treated as perfect, with zero heat transfer capability. This assumption is typical for stress analyses in similar components.

Pressure stresses are calculated at a stress-free temperature of 70°F and do not include any thermal stress effects.

The density and Poissons ratio are assumed temperature independent.

The W/CE insurge/outsurge transient Group 3 is used and scaled based on the methodology outlined in Reference [4] to obtain the B&W insurge/outsurge transients.

5.0 CALCULATIONS 5.1 Finite Element Model A finite element model of the PZR surge nozzle, bottom head and lower shell is developed using the ANSYS finite element analysis software package [1], with dimensions shown in Figure 1.

Because of the axisymmetric nature of this configuration, a 3-D model is constructed using 3-D structural solid, SOLID185, elements. The thermal equivalent element for the thermal transient analyses is SOLID70. The weld locations are shown in Figure 2 and are based on Reference [6]. The constructed model is shown in Figure 3.

File No.: 2100561.302 Revision: 1 Page 6 of 38 F0306-01R4 5.1.1 Material Properties The Oconee pressurizer nozzles are fabricated using SA-508, Grade 1, Class 1 (Carbon Steel), while the pressurizer heads and vessels are fabricated from SA-212, Grade B (Unit 1) or SA-516, Grade 70 (Unit 2 and 3) (Carbon Steel) [2, Table 1]. A typical stainless steel (SA-240, Type 304) is assumed for all cladding material.

The material properties were obtained from the relevant tables in the 2013 Edition of ASME Code,Section II, Part D [8]. Temperature dependent material properties used in the finite element analysis are listed in Table 1 and Table 2.

5.2 Pressure / Thermal Stress Analysis 5.2.1 Unit Internal Pressure Loading Analysis A unit internal pressure of 1,000 psi is applied to the interior surfaces of the model. The resulting stresses will be scaled to the appropriate plant pressure conditions for subsequent fracture mechanics evaluations. An induced end-cap load is applied to the free end of the surge nozzle in the form of tensile axial pressures, as calculated below.

Pnozzle-cap =

2 22 =

10008.752 11.528.752 = 1,375 psi

where, Pnozzle-cap =

End cap pressure on nozzle free end (psi)

P

=

Internal pressure (psi)

ID

=

Inside diameter of nozzle free end (in)

OD

=

Outside diameter of nozzle free end (in)

Symmetric boundary conditions are applied to the axial and circumferential free ends of the pressurizer shell while axial displacement couples are applied to the free end of the surge nozzle. The applied pressure load and boundary conditions for this case are shown in Figure 4.

5.2.2 Thermal Heat Transfer Analyses Thermal transient parameters were developed in Reference [4] for use in this analysis. The thermal transients listed in Table 3 and Table 4, are applied to the interior surface nodes of the nozzle and head.

The heat transfer coefficients for the inside surface of the surge nozzle and shell are also listed in Table 3 and Table 4.

Per Section 4.0, no heat transfer coefficients or temperatures are applied to the assumed insulated outside surfaces. Figure 5 shows representative plots of the thermal loads applied for Heatup/Cooldown transient.

For the insurge and outsurge transients defined in Table 4 a pool height, temperature and heat transfer coefficient are also specified. These represents the pool of colder water that builds up in the pressurizer during the insurge and then its draining during the outsurge.

Examining Table 4 shows that the pool heat transfer coefficient is relatively low (50 Btu/hr-ft²-°F), due to the mixing process of the surge inflow with the relatively stagnant pressurizer fluid. However, in order to introduce added conservativism, the pool heat transfer will be left at the higher heat transfer value at insurge initiation for the entire length of the insurge / outsurge event. Figure 6 shows representative plots of the thermal loads applied for Insurge/Outsurge transient Group 3.

File No.: 2100561.302 Revision: 1 Page 7 of 38 F0306-01R4 An additional time of 3,600 seconds is added to the end of each transient to ensure that any lagging peak stresses are captured, followed by a steady state load step (at an arbitrary 400 seconds after the 3,600 seconds of additional time).

5.2.3 Thermal Stress Analyses Symmetric boundary conditions are applied to the axial and circumferential free ends of the pressurizer shell, while axial displacement couples are applied on the free end of the surge nozzle. The stress-free reference temperature for the thermal strain calculation is assumed to be 70°F. Figure 7 shows a plot of the boundary conditions applied for the thermal stress analyses.

6.0 RESULTS OF ANALYSIS A finite element model of a typical B&W pressurizer surge nozzle, bottom head, lower shell and heater bundle shell thickness region has been developed. The stress results due to thermal transients and unit internal pressure have been run and are stored to be used for future fracture mechanics evaluations.

Example hoop stress and axial stress contour plots for unit internal pressure are shown in Figure 8.

Representative temperature contour and stress contour plots for the Heatup/Cooldown are shown in Figure 9 and Figure 10, respectively. The instance shown in Figure 9 and Figure 10 is when the maximum stress intensity occurs. Representative temperature contour and stress contour plots for the Insurge/Outsurge Transient Group 3 at 900 second are shown in Figure 11 and Figure 12, respectively.

Figure 13 shows the path locations where stresses are extracted. Path 1 through Path 3 are chosen as representative of the nozzle-to-lower head weld. Path 4 though Path 7 are chosen as a representative of the shell-to-lower head weld and heater bundle shell thickness region-to-lower head weld. Path 8 through Path 11 are chosen as a representative of the lower shell-to-upper shell weld and heater bundle shell thickness region-to-upper shell weld. Path 12 is representative of the heater bundle shell thickness region-to-lower shell axial weld and Path 13 is representative of the heater bundle shell thickness region middle circumferential weld. The input and output files used in this evaluation are listed in Appendix A.

All stresses are extracted in a cylindrical coordinate system, about the pressurizer surge nozzle.

Through-wall stress distributions for Path 1 through Path 13 are shown in Figure 14 and Figure 26, respectively.

7.0 REFERENCES

1. ANSYS Mechanical APDL (UP20170403) and Workbench (March 31, 2017), Release 18.1, SAS IP, Inc.

2. SI Calculation No. ONS-15Q-312, Rev. 1, Finite Element Model of Pressurizer Surge Nozzle with Weld Overlay Repair Per Design Dimensions.

3. Babcock and Wilcox Drawing No. 25476, Rev. 7, Pressurizer General Arrangement, SI File No.

ANO-39Q-212.

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

5. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2017 Edition.

6. Oconee Drawing No. ISI-OCNI-002, Rev. 2, Pressurizer Weld Outline, SI File No.

2100561.205.

File No.: 2100561.302 Revision: 1 Page 8 of 38 F0306-01R4

7. Oconee Drawing No. OM 201-1878, Rev. 6, Radiographic Outline, SI File No. 2100561.205.

8. ASME Boiler and Pressure Vessel Code,Section II Materials, Part D - Properties, 2013 Edition.

9. Letter from Austin Keller (Duke Energy) to Scott Chesworth (SI) on August 3, 2022,

Subject:

Duke Energy Inputs for SIA Calculation 2100561.302, SI File No. 2100561.205.

File No.: 2100561.302 Revision: 1 Page 9 of 38 F0306-01R4 Table 1. Material Properties for Carbon Steel, C>0.3% (SA-508, Class 1 or SA-516, Grade 70)

Temperature

(°F)

Modulus of Elasticity (E)

(106 psi)

Coefficient of Thermal Expansion ()

(10-6 in/in/°F)

Thermal Conductivity (K)

(10-4 BTU/in-s-°F)

Specific Heat (C)(4)

(BTU/lb-°F) 70 29.2 6.4 8.08 0.103 100 29.1(1) 6.5 8.03 0.106 150 28.8(1) 6.6 7.92 0.110 200 28.6 6.7 7.80 0.114 250 28.4(1) 6.8 7.64 0.117 300 28.1 6.9 7.48 0.119 350 27.9(1) 7.0 7.31 0.122 400 27.7 7.1 7.15 0.124 450 27.4(1) 7.2 6.97 0.126 500 27.1 7.3 6.81 0.128 550 26.8(1) 7.3 6.64 0.131 600 26.4 7.4 6.48 0.134 650 25.9(1) 7.5 6.32 0.136 700 25.3 7.6 6.16 0.140 Notes:

1. Linearly interpolated.

2. Density () = 0.280 lb/in3 [8, Table PRD], assumed temperature independent.

3. Poissons Ratio () = 0.3 [8, Table PRD], assumed temperature independent.

4. Calculated per Note 1 of Table TCD [8].

5. SA-508, Class 1 has been reclassified as SA-508, Grade 1, Class 1 in Reference [8].

File No.: 2100561.302 Revision: 1 Page 10 of 38 F0306-01R4 Table 2. Material Properties for Stainless Steel (SA-240, Type 304)

Temperature

(°F)

Modulus of Elasticity (E)

(106 psi)

Coefficient of Thermal Expansion ()

(10-6 in/in/°F)

Thermal Conductivity (K)

(10-4 BTU/in-s-°F)

Specific Heat (C)(4)

(BTU/lb-°F) 70 28.3 8.5 1.99 0.114 100 28.1(1) 8.6 2.01 0.114 150 27.8(1) 8.8 2.08 0.117 200 27.5 8.9 2.15 0.119 250 27.3(1) 9.1 2.22 0.121 300 27.0 9.2 2.27 0.122 350 26.7(1) 9.4 2.34 0.124 400 26.4 9.5 2.41 0.126 450 26.2(1) 9.6 2.45 0.127 500 25.9 9.7 2.52 0.129 550 25.6(1) 9.8 2.57 0.129 600 25.3 9.9 2.62 0.130 650 25.1(1) 9.9 2.69 0.131 700 24.8 10.0 2.73 0.132 Notes:

1. Linearly interpolated

2. Density () = 0.290 lb/in3 [8, Table PRD], assumed temperature independent.

3. Poissons Ratio () = 0.31 [8, Table PRD], assumed temperature independent.

4. Calculated per Note 1 of Table TCD [8].

File No.: 2100561.302 Revision: 1 Page 11 of 38 F0306-01R4 Table 3. Thermal Transients for Pressurizer Surge Nozzle Transient

Time, sec

Tpzr,

°F

Tnoz,

°F

Press, psig Heat Transfer Coefficients, BTU/hr-ft2-°F Surge Nozzle Thermal Sleeve Region in Surge Nozzle (2)

Bottom Head

& Lower Shell Heatup /

Cooldown 0

70 70 0

300 100 300 10494 653 653 2300 39294 653 653 2300 49788 (3) 70 70 0

Loss of Load 0

653 653 Max. 2710 Min. 1685 15000 150 300 10 672 644.5 20 672 617 40 672 614.7 71 610 549.2 240 594 514.0 600 560 480 2000 560 560 Notes:

1. Above table is reproduced from Table 5-6 of Reference [4].

2. The B&W surge nozzle does not include a thermal sleeve and the heat transfer coefficient for the Thermal Sleeve Region in the Surge Nozzle is ignored.

3. Note that an end time of 51,048 seconds was run instead. This results in a cooldown rate of 178.45°F/hr vs. a design of 200°F/hr. The cooldown rate of 178.45°F/hr. used in the analysis (vs. a design of 200°F/hr.) only affects the cooldown ramp; the heatup ramp is not impacted. The change in cooldown rate is expected to have minimal impact on the stresses.

File No.: 2100561.302 Revision: 1 Page 12 of 38 F0306-01R4 Table 4. Insurge/Outsurge Transients for Pressurizer Surge Nozzle Transient Group Time sec Pzr Temp oF Pzr Press psig Pool Temp oF Pool Height ft Fluid Temp at Pzr Nozzle oF HPzr Btu/hr-ft2-°F HPool Btu/hr-ft2-°F HNoz Btu/hr-ft2-°F HTS Btu/hr-ft2-°F (3)

Group 3 0

550 1000 550 0

550 300 200 300 100 900 550 1000 390 12 220 300 200 300 100 5200 550 1000 400 12 220 300 50 (2) 300 100 5201 550 1000 400 12 400 400 50 (2) 300 100 6200 550 1000 400 0

400 400 50 (2) 300 100 6201 550 1000 550 0

550 400 50 (2) 300 100 Notes:

1. Above table is reproduced from Table 5-9 of Reference [4].

2. Pool heat transfer coefficient conservatively increased to 200 Btu/hr-ft²-°F.

3. The B&W surge nozzle does not include a thermal sleeve and the heat transfer coefficient for the Thermal Sleeve Region in the Surge Nozzle is ignored.

File No.: 2100561.302 Revision: 1 Page 13 of 38 F0306-01R4 Figure 1. Modeled Dimensions (Dimensions Based on Figure 1 of Reference [2], Reference [3] and Reference [6].)

File No.: 2100561.302 Revision: 1 Page 14 of 38 F0306-01R4 Figure 2. Weld Locations (Weld Location based on Reference [6].)

File No.: 2100561.302 Revision: 1 Page 15 of 38 F0306-01R4 Figure 3. 3-D Finite Element Model

File No.: 2100561.302 Revision: 1 Page 16 of 38 F0306-01R4 Figure 4. Applied Boundary Conditions and Unit Internal Pressure (Units for pressure is psi.)

File No.: 2100561.302 Revision: 1 Page 17 of 38 F0306-01R4 Heat Transfer Coefficient Bulk Temperature Figure 5. Example of Applied Thermal Boundary Conditions for Thermal Transient Analyses Heatup/Cooldown transient shown; loads applied at time = 51,048 seconds.

(Units for HTC is BTU/sec-in2-°F, TBULK is °F.)

File No.: 2100561.302 Revision: 1 Page 18 of 38 F0306-01R4 Heat Transfer Coefficient Bulk Temperature Figure 6. Example of Applied Thermal Boundary Conditions for Insurge/Outsurge Transient Analyses Insurge/Outsurge transient Group 3 shown, loads applied at time = 900 seconds.

(Units for HTC is BTU/sec-in2-°F, TBULK is °F.)

File No.: 2100561.302 Revision: 1 Page 19 of 38 F0306-01R4 Figure 7. Applied Mechanical Boundary Conditions for Thermal Stress Analyses

File No.: 2100561.302 Revision: 1 Page 20 of 38 F0306-01R4 X-direction Stress (Radial Stress to the Nozzle) Y-direction Stress (Axial Stress to the Nozzle)

Z-direction Stress (Hoop Stress to the Nozzle)

Figure 8. Stress Contours Due to Unit Internal Pressure (Units for stress is psi.)

File No.: 2100561.302 Revision: 1 Page 21 of 38 F0306-01R4 Figure 9. Temperature Contour Heatup/Cooldown Transient (Time = 10,494 seconds)

(Units for temperature is °F.)

File No.: 2100561.302 Revision: 1 Page 22 of 38 F0306-01R4 X-direction Stress (Radial Stress to the Nozzle) Y-direction Stress (Axial Stress to the Nozzle)

Z-direction Stress (Hoop Stress to the Nozzle)

Figure 10. Stress Contours of Heatup/Cooldown Transient (Time = 10,494 seconds)

(Units for stress is psi.)

File No.: 2100561.302 Revision: 1 Page 23 of 38 F0306-01R4 Figure 11. Temperature Contour Insurge/Outsurge Group 3 Transient (Time = 900 seconds)

(Units for temperature is °F.)

File No.: 2100561.302 Revision: 1 Page 24 of 38 F0306-01R4 X-direction Stress (Radial Stress to the Nozzle) Y-direction Stress (Axial Stress to the Nozzle)

Z-direction Stress (Hoop Stress to the Nozzle)

Figure 12. Stress Contours of Insurge/Outsurge Group 3 Transient (Time = 900 seconds)

(Units for stress is psi.)

File No.: 2100561.302 Revision: 1 Page 25 of 38 F0306-01R4 Figure 13. Path Locations

File No.: 2100561.302 Revision: 1 Page 26 of 38 F0306-01R4 Figure 14. Through-Wall Stress Distribution at Path 1

File No.: 2100561.302 Revision: 1 Page 27 of 38 F0306-01R4 Figure 15. Through-Wall Stress Distribution at Path 2

File No.: 2100561.302 Revision: 1 Page 28 of 38 F0306-01R4 Figure 16. Through-Wall Stress Distribution at Path 3

File No.: 2100561.302 Revision: 1 Page 29 of 38 F0306-01R4 Figure 17. Through-Wall Stress Distribution at Path 4

File No.: 2100561.302 Revision: 1 Page 30 of 38 F0306-01R4 Figure 18. Through-Wall Stress Distribution at Path 5

File No.: 2100561.302 Revision: 1 Page 31 of 38 F0306-01R4 Figure 19. Through-Wall Stress Distribution at Path 6

File No.: 2100561.302 Revision: 1 Page 32 of 38 F0306-01R4 Figure 20. Through-Wall Stress Distribution at Path 7

File No.: 2100561.302 Revision: 1 Page 33 of 38 F0306-01R4 Figure 21. Through-Wall Stress Distribution at Path 8

File No.: 2100561.302 Revision: 1 Page 34 of 38 F0306-01R4 Figure 22. Through-Wall Stress Distribution at Path 9

File No.: 2100561.302 Revision: 1 Page 35 of 38 F0306-01R4 Figure 23. Through-Wall Stress Distribution at Path 10

File No.: 2100561.302 Revision: 1 Page 36 of 38 F0306-01R4 Figure 24. Through-Wall Stress Distribution at Path 11

File No.: 2100561.302 Revision: 1 Page 37 of 38 F0306-01R4 Figure 25. Through-Wall Stress Distribution at Path 12

File No.: 2100561.302 Revision: 1 Page 38 of 38 F0306-01R4 Figure 26. Through-Wall Stress Distribution at Path 13

File No.: 2100561.302 Revision: 1 Page A-1 of A-2 F0306-01R4 COMPUTER FILES LISTING

File No.: 2100561.302 Revision: 1 Page A-2 of A-2 F0306-01R4 File Name Description BW-Surge-Geom.INP Input file to construct model for a B&W PWR pressurizer surge nozzle, bottom head and lower shell.

PRESS.INP Unit internal pressure input file HUCD.INP Plant Heatup/Cooldown thermal analysis file LOL.INP Loss of Load thermal analysis file IO_GP3.INP Insurge/Outsurge Group 3 thermal analysis file STRESS.INP Input file for stress analyses for thermal transients CMNTR.mac ANSYS macro used to develop temperature load files during stress analysis.

$$_mntr.inp Thermal analysis load step input file for thermal transients, $$ =

HUCD for Heatup/Cooldown, LOL for Loss of Load, and IO_GP3 for Insurge/Outsurge Group 3 Post.INP Input file for post-processing GenStress.mac ANSYS Macro for stress extraction GETPATH.TXT Input file for defining stress Path 1 through Path 13

$$_MAP_P%.CSV Output files containing mapped stresses, $$ = STR_HUCD for Heatup/Cooldown, STR_LOL for Loss of Load, STR_IO_GP3 for Insurge/Outsurge Group 3, and PRESS for unit internal pressure loading, % = Paths 1 - 13 Results.xlsx Excel file to create Figure 14 through Figure 26.

L-24-233 SI Calculation No. 2100561.303, Deterministic and Probabilistic Fracture Mechanics of Oconee Units 1, 2, and 3 Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head, Revision 2 (40 pages follow)

CALCULATION PACKAGE File No.: 2100561.303 Project No.: 2100561 Quality Program Type:

Nuclear Commercial PROJECT NAME:

Duke PWR Fleet Steam Generator & Pressurizer Inspection Optimization Relief Requests CONTRACT NO.:

GSA # 03021365 CLIENT:

Duke Energy Corporation PLANT:

Oconee Nuclear Station, Units 1, 2 and 3 CALCULATION TITLE:

Deterministic and Probabilistic Fracture Mechanics Analyses of Oconee Units 1, 2 and 3 Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head Document Revision Affected Pages Revision Description Project Manager Approval Signature & Date Preparer(s) &

Checker(s)

Signatures & Date 0

1 - 38 A A-2 Initial Issue Scott Chesworth 5/27/2022 Dilip Dedhia 5/27/2022 Nathaniel G. Cofie 5/27/2022 1

11 Revised Reference [5]

information; removed Proprietary notation from all locations Scott Chesworth 6/2/2022 Scott Chesworth 6/2/2022 Nathaniel G. Cofie 6/2/2022 2

6, 11 Corrected Reference [1];

added clarifying wording to Sections 2.2.3.1 and 4.0 Scott Chesworth 8/8/2022 Scott Chesworth 8/8/2022 Nathaniel G. Cofie 8/8/2022

File No.: 2100561.303 Revision: 2 Page 2 of 38 F0306-01R4 Table of Contents 1.0 OBJECTIVE................................................................................................................ 5 2.0 DFM EVALUATION..................................................................................................... 5 2.1 Technical Approach......................................................................................... 5 2.2 Design Inputs.................................................................................................. 5 2.3 Results of Deterministic Fracture Mechanics Evaluation................................ 9 3.0 PFM EVALUATION..................................................................................................... 9 3.1 Technical Approach......................................................................................... 9 3.2 Design Inputs.................................................................................................. 9 3.3 Inspection Coverage..................................................................................... 10 3.4 Results of PFM Evaluation............................................................................ 10

4.0 CONCLUSION

S........................................................................................................ 10

5.0 REFERENCES

.......................................................................................................... 11 COMPUTER FILES LISTING........................................................................ A-1

File No.: 2100561.303 Revision: 2 Page 3 of 38 F0306-01R4 List of Tables Table 1: Summary of DFM Design Inputs............................................................................. 13 Table 2: ONS 1/2/3 Plant Specific Transient Cycles Used in the DM and PFM Evaluations 13 Table 3: Summary of ONS 1/2/3 Plant Specific Insurge/Outsurge Temperature Differences and Numbers of Cycles Used in the DFM Evaluations(1)............................................ 14 Table 4: Results of the DFM Evaluation................................................................................ 15 Table 5: PFM Inputs for ONS 1/2/3 Inspection Scenario...................................................... 16 Table 6: Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for Oconee Units 1, 2 and 3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70)

............................................................................................................................ 17 Table 7: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness............................................. 18 Table 8: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Stress................................................................... 19 Table 9: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness and Stress........................... 20 Table 10: Sensitivity Study for ISI Examination Coverage for ONS 1/2/3 B3.110 Welds...... 21

File No.: 2100561.303 Revision: 2 Page 4 of 38 F0306-01R4 List of Figures Figure 1. Modeled Dimensions (From Reference [2]).......................................................... 22 Figure 2: Path Locations (From Reference [2]).................................................................... 23 Figure 3 Through-Wall Stress Distribution for Path P1 [2]................................................... 24 Figure 4. Through-Wall Stress Distribution for Path P2 [2].................................................. 25 Figure 5. Through-Wall Stress Distribution for Path P3 [2].................................................. 26 Figure 6. Through-Wall Stress Distribution for Path P4 [2].................................................. 27 Figure 7. Through-Wall Stress Distribution for Path P5 [2].................................................. 28 Figure 8. Through-Wall Stress Distribution for Path P6 [2].................................................. 29 Figure 9. Through-Wall Stress Distribution for Path P7 [2].................................................. 30 Figure 10. Through-Wall Stress Distribution for Path P8 [2]................................................ 31 Figure 11. Through-Wall Stress Distribution for Path P9 [2]................................................ 32 Figure 12. Through-Wall Stress Distribution for Path P10 [2].............................................. 33 Figure 13. Through-Wall Stress Distribution for Path P11 [2].............................................. 34 Figure 14. Through-Wall Stress Distribution for Path P12 [2].............................................. 35 Figure 15. Through-Wall Stress Distribution for Path P13 [2].............................................. 36 Figure 16. Weld Residual Stress Distribution....................................................................... 37 Figure 17. Semi-Elliptical Axial Crack in a Cylinder Model.................................................. 37 Figure 18. Semi-Elliptical Circumferential Crack in a Cylinder Model.................................. 37 Figure 19. The Effect of Temperature on the Fracture Toughness, JIc, of SA-516 Grade 70 Steel

[11]...................................................................................................................... 38

File No.: 2100561.303 Revision: 2 Page 5 of 38 F0306-01R4 1.0 OBJECTIVE EPRI Report 3002015906 [1] developed the technical basis for optimizing the examination of PWR pressurizer shell and nozzle weld components. The pressurizer configuration in the Reference [1]

report is representative of Westinghouse and CE designs. The configuration of the B&W design pressurizers at Oconee Units 1, 2 and 3 (ONS 1/2/3) is considerably different from the Westinghouse/CE designs in terms of both geometry and materials; therefore, an evaluation is required to address the B&W pressurizer configuration. To this end, stress analyses of the ONS 1/2/3 design were performed in Reference [2] for the pressurizer bottom head and in Reference [3] for the pressurizer top head.

The objective of this calculation is to perform deterministic fracture mechanics (DFM) and probabilistic fracture mechanics (PFM) analyses of the B&W pressurizer components using the results of the stress analysis in Reference [2] and other relevant design inputs. The DFM evaluation will determine how long a postulated flaw will take to reach the ASME Code allowable flaw size, while the PFM evaluation will determine the probabilities of failure (leak and rupture) at the component locations.

The ASME Code,Section XI item numbers associated with the ONS 1/2/3 pressurizer components are:

Item No. B2.11 - Pressurizer, shell-to-head welds, circumferential Item No. B2.12 - Pressurizer, shell-to-head welds, longitudinal Item No. B3.110 - Pressurizer, nozzle-to-vessel welds 2.0 DFM EVALUATION 2.1 Technical Approach The technical approach used in the DFM evaluation is to postulate an initial flaw size equivalent to the relevant ASME Code,Section XI acceptance standard [4]. The ASME Code,Section XI fatigue crack growth (FCG) law, with the through-wall stress distributions from References [2, 3] and appropriate fracture mechanics models, is then used to determine the length of time for the postulated initial flaw to grow to a depth of 80% of the wall thickness (assumed to equate to leakage in this evaluation) or the depth at which the allowable toughness (KIC reduced by a structural factor of 2.0 for primary stresses and 1.0 for secondary stresses) is reached, whichever is less.

2.2 Design Inputs The design inputs used in the DFM evaluation are summarized in Table 1 and discussed in the following sections.

2.2.1 Geometry The geometries of the B&W pressurizer components considered in the evaluation are presented in Figure 1 [2]. The specific components of interest are the nozzle-to-head welds and head-to-shell welds for the pressurizer bottom head (Note: comparison of the stresses in References [2, 3] indicate the stresses of the bottom head in Reference [2] are controlling). The dimensions at these welds are provided in Figure 1. Figure 2 [2] shows the stress paths where these welds are located.

File No.: 2100561.303 Revision: 2 Page 6 of 38 F0306-01R4 2.2.2 Initial Crack Size and Shape For all components, an initial crack size of 5.2% of the wall thickness (which corresponds to the most conservative flaw acceptance standard for these components from Table IWB-3510-1 of ASME Code,Section XI [4]) was used in the DFM evaluation. This initial crack depth is the maximum value from these two tables with an associated crack aspect ratio (half crack length-to-crack depth) of 1.0. This crack shape results in the most conservative initial stress intensity factor (K) at the deepest point of the crack. The aspect ratio is then subsequently allowed to vary during the crack growth process.

2.2.3 Applied Stresses 2.2.3.1 Operating Transient Stresses Comparison of the stresses in References [2, 3] indicate the stresses of the bottom head in Reference

[2] are controlling and therefore will be used in this evaluation. The applied stresses consist of through-wall stresses due to pressure and the thermal transients described in Reference [2]. Typical through-wall stress distributions for stress paths used in the evaluation from Reference [2] are reproduced in Figures 3 to 15. Figure 2 shows the stress paths where these stresses were extracted. Plant-specific number of cycles for ONS 1/2/3 general transients [5] and insurge/outsurge transients [6] shown in Tables 2 and 3 (respectively) were used in the evaluations. It should be noted that for the PFM evaluation, the conservative number of cycles associated with transients defined in Reference [2] was used in lieu of the plant-specific transients.

2.2.3.2 Weld Residual Stresses Pressure vessel welds typically receive post-weld heat treatment (PWHT) to reduce the effects of weld residual stresses. In this evaluation, weld residual stresses remaining after PWHT were characterized in the form of a cosine distribution with a peak stress of 8 ksi [7] as shown in Figure 16, consistent with what was used in Reference [1].

2.2.4 Fracture Mechanics Models In this evaluation, all pre-existing flaws were conservatively assumed to be surface flaws. Two different fracture mechanics models were used for axial and circumferential flaws. For an axial flaw, the stress intensity factor (K) solution for an internal, semi-elliptical crack from API-579/ASME-FFS-1 [8] was used.

This model is shown in Figure 17. The aspect ratio (a/c) is allowed to vary during crack growth.

For a circumferential flaw, the K solution for an internal, semi-elliptical crack from API-579/ASME-FFS-1

[8] was used. This model is shown in Figure 18. The aspect ratio (a/c) was allowed to vary during crack growth. These models are consistent with those used in Reference [1].

These fracture mechanics models were incorporated into an SI-developed software, SI-TIFFANY [9],

that determines the K distribution due to through-wall stress profiles for both circumferential and axial cracks. The outputs of SI-TIFFANY are the maximum and minimum K distributions as well as the K distribution for each transient.

File No.: 2100561.303 Revision: 2 Page 7 of 38 F0306-01R4 2.2.5 Fracture Toughness The materials under consideration are low carbon ferritic steels (SA-516 Grade 70 and SA-212). The fracture toughness provided in ASME Code,Section XI, Appendix A [4] only applies to low alloy ferritic steels such as SA-533 Grade B Class 1, SA-508 Class 2, and SA-508 Class 3 (typically used in the fabrication of Westinghouse and CE pressurizers) and is therefore not applicable to the B&W pressurizer design. The fracture toughness for low carbon ferritic steel piping components is provided in Appendix C of ASME Code,Section XI. Though the pressurizers are not piping components, guidance will be taken from Appendix C in addition to information in the open literature in determining a reasonable lower bound fracture toughness for use in this evaluation. Lower bound values of fracture toughness for ferritic piping are provided in ASME Code,Section XI, Appendix C for both circumferential and axial flaws.

Table C-8321-1 of Appendix C provides fracture toughness values for two categories of ferritic materials operating in the upper shelf region for circumferential flaws. The first material category includes seamless or welded wrought ferritic steel pipe and pipe fittings that have a material yield strength not greater than 40 ksi (280 MPa) and welds made with E7015, E7016, and E7018 electrodes in the as-welded or PWHT condition. The fracture toughness (JIc) value for circumferential flaws for this material category is 600 in.-lb/in.2 (105 kN/m) at temperatures equal to or greater than the upper shelf temperature. The second material category is for all other ferritic shielded metal arc and submerged arc weld with a specified minimum tensile strength not greater than 80 ksi (550 MPa) in the as-welded or post-weld heat treated condition for circumferential flaws. The JIc value for this material category is 350 in.-lb/in.2 (61 kN/m) at temperatures equal to or greater than the upper shelf temperature.

Similarly, the JIc value for axial flaws for from Table C-8322-1 of Appendix C for base metals and weldments is 300 in.-lb/in.2 (53 kN/m) at temperatures equal to or greater than the upper shelf temperature. However, per the technical basis document for ferritic piping [10], this value of fracture toughness is associated with small diameter piping (4-inch and 6-inch NPS) for SA-106 Grade B material and was determined in the C-L direction for through-wall flaws. Since the pressurizer components are much larger in diameter and part-wall flaws are considered in this evaluation, this fracture toughness is not considered appropriate for use in the evaluation of the pressurizer components. Therefore, the lower bound fracture toughness value in the circumferential direction was used in determining the flaw acceptance criteria for all flaws in the pressurizer components in this evaluation.

From Table C-8321-1 of Appendix C, circumferential flaws in ferritic steel base metals and weldments that have thicknesses equal to or greater than one inch and an operating temperature greater than 52oF (11°C) are in the upper shelf region. The thickness of one inch or greater is consistent with the dimensions of the pressurizer shown in Figure 1.

Using the relation from Table C-8321 of the ASME Code, the value of the plane strain fracture toughness (KIc) can be determined from:

J1c = 1000 (KIc)2 / E Eq. 1 where E = E / (1 - 2)

E = modulus of elasticity = 29,200 ksi at 70°F (from Table 5-2)

File No.: 2100561.303 Revision: 2 Page 8 of 38 F0306-01R4

= Poisson ratio = 0.3 Using Equation 1, the values of JIc of 350 in.-lb/in.2 translates into KIc value of 106 ksiin. This lower value bound value of KIc will be used in this evaluation.

The fracture toughness determined above is applicable for piping components. The pressurizer under consideration in this study is comprised of a cylindrical shell and nozzle. As such, the reasonableness of applying a fracture toughness value for piping to the components is investigated for SA-516 Grade 70 and SA-212.

The fracture toughness for SA-516 Grade 70, typical of what is used to fabricate pressurizer shells has been investigated by several researchers [11, 12, 13]. A plot of the fracture toughness, JIc, versus temperature for SA-516, Grade 70 from Reference [11] is shown in Figure 19. Per Reference [11], the JKe data (black solid data) is only valid in the temperature range of -130oC to -160oC. As indicated in this figure, JIc remains constant at temperatures above 0°C (32°F) at a value of 120 kN/m (685 in.-

lb/in2). This upper shelf toughness is greater than the JIc value of 350 in.-lb/in.2 established above.

Studies performed in References [12, 13] determined the average value of KIc at room temperature for SA-516 Grade 70 steel base material and weldments to be 129 MPam (117 ksiin), which is greater than the 106 ksiin established above. In addition, data presented in Reference [11] for various ferritic steels show that the fracture toughness (JIc) of SA-516 Grade 70 material at 550oF is at least 800 in.-

lb/in2. Therefore, the lower bound upper shelf fracture toughness of 106 ksiin derived from Appendix C of Section XI of the ASME Code for circumferential flaws in piping is a conservative lower bound for use in this evaluation for the SA-106 pressurizer material at ONS 1/2/3.

SA-212 is an older version of SA-516, Grade 70. It first occurred in the 1940 Edition of ASME Code,Section II. In the 1968 edition of Section II, the SA-212 Specification was deleted and was replaced with two specifications, SA-515 (Specification for Pressure Vessel Plates, Carbon Steel, for Intermediate and Higher Temperature Service) and the SA-212 steel plate melted to fine grain practice was replaced with SA-516 (Specification for Pressure Vessel Plates, Carbon Steel, for Moderate and Lower Temperature Service). SA-212 was the material of choice for fabrication of railroad tank cars. Data presented n Table 4-11 of Reference [14] indicate that the upper shelf energy for SA-212 range from 45 to 68 ft lbs.

in the longitudinal direction and 30 to 34 ft lbs. in the transverse direction. Data from Table A1 of Reference [15] also show that the minimum upper shelf energy at temperatures greater than 212oF is 71.4 ft-lbs. and from Table A4, it is 45.1 ft-lbs. Using Equation 3.2 from Reference [16]:

(KIc)2/E = 2(CVN)3/2 (psi-in., ft-lb.)

Eq. 2 where:

KIc = fracture toughness E = modulus of elasticity = 29,200 ksi at 70°F CVN = energy absorbed Using a reasonable lower bound value of CVN of 45 ft-lbs. from the data in References [14, 15] and the KIc-CVN correlation from Equation 2 results in KIc of 132.3 ksiin. Therefore, the lower bound upper shelf fracture toughness of 106 ksiin derived from Appendix C of Section XI of the ASME Code for

File No.: 2100561.303 Revision: 2 Page 9 of 38 F0306-01R4 circumferential flaws in piping is a conservative lower bound for use in this evaluation in application to SA-212 pressurizer material at ONS 1/2/3.

2.2.6 Fatigue Crack Growth Law The FCG law for ferritic steels, as defined in ASME Code,Section XI, Appendix A, Paragraph A-4300

[4], was used in the evaluation.

2.3 Results of Deterministic Fracture Mechanics Evaluation The results of the DFM evaluation are summarized in Table 2. The table shows that the periods required for hypothetical postulated flaws to reach the allowable fracture toughness or 80% of thickness are very long, which indicates that all the evaluated components are very flaw tolerant. Because the DFM evaluation considered hypothetical postulated flaws, structural factors of 2.0 on primary loads and 1.0 on secondary loads, consistent with ASME Code,Section XI, Appendix G, were applied. Hence, a structural factor of 2.0 was applied to the applied K due to pressure stress and a structural factor of 1.0 was applied to the thermal and residual stresses. The resulting K with the structural factors were compared to the allowable fracture toughness of 106 ksiin to determine the allowable operating period.

Table 5 shows that the number of years to reach the fracture toughness with a structural factor of 2.0 on primary stresses and 1.0 on secondary stresses for all locations. As seen in this table, it takes a minimum of 208 years at the limiting location (Case ID PRSHC-BW-4C) to reach the fracture toughness.

3.0 PFM EVALUATION 3.1 Technical Approach The PFM evaluation was performed consistent with the evaluation presented in the EPRI Report [1].

PROMISE, Version 2.0 [17] was used to perform the PFM evaluations. The evaluation considered the ONS 1/2/3 plant-specific inspection history.

3.2 Design Inputs The design inputs used for the PFM evaluation are shown in Table 5 for ONS 1/2/3 plant specific inspection history and are consistent with those used in Reference 1. ONS 1/2/3 has performed pre-service (PSI) inspection at year zero followed by four successive 10-year in-service inspections (ISI).

This is to be followed by 30-year inspection deferral which is being considered in the Relief Request being developed by Duke Energy. This plant specific history was considered in the PFM evaluation.

Stress and fracture toughness which were identified as the key variables in the PFM evaluation in Reference [1]. As such, three sensitivity studies were performed as part of the PFM as follows:

1.

The fracture toughness was decreased to determine the minimum fracture toughness that will meet the acceptance criteria of 1.0x10-6.

2.

The stresses were increased to determine the maximum stress multiplier that will meet the acceptance criteria of 1.0x10-6.

3.

A sensitivity study of the combined effects of the fracture toughness and stress.

File No.: 2100561.303 Revision: 2 Page 10 of 38 F0306-01R4 3.3 Inspection Coverage Inspection coverage for all the welds under consideration for ONS 1/2/3 is provided in Reference [18].

For Item Nos. B2.11 and B2.12, the inspection coverage is greater than 90% (essentially 100%) for all welds. However, for Items B3.110, some welds have limited coverage. The minimum coverage for Unit 1 is 25.8%, for Unit 2 is 25.2 and for Unit 3 is 30.0%. A sensitivity study is performed with the limiting minimum coverage of 25.2% for Unit 2. Evaluations were performed using this limiting coverage to determine the probabilities of rupture and leakage for the plant-specific inspection scenarios of (PSI+10+10+30+40+70) using the same input parameters as in Table 6. For comparison, evaluations were also performed for the current ASME Code,Section XI mandated 10-year inspection interval of (PSI+10+20+30+40+50+60+70).

3.4 Results of PFM Evaluation The results of the PFM evaluation are presented in Table 6 for ONS 1/2/3 plant specific inspection history. As shown in this table, the probabilities of rupture and leakage are all below the acceptance criteria of 1.0x10-6 after 80 years of plant operation by three orders of magnitude.

The results of the sensitivity studies are presented in Tables 7 through 9. From Table 7, the fracture toughness can be as low as 72 ksiin before the acceptance criterion of 1.0x10-6 is reached after 80 years of operation. From Table 8, a stress multiplier of 1.4 can be applied to all the stresses considered in the evaluation before the acceptance criterion is reached. Table 9 shows that by applying a stress multiplier of 1.1 and reducing the fracture toughness to 80 ksiin, the probabilities of rupture and leakage are all below the acceptance criterion of 1.0x10-6 after 80 years of plant operation.

The results of the sensitivity study on coverage are presented in Table 10. As shown in Table 10, considering the most limiting coverage for Item No. B3.110 and the ONS 1/2/3 PSI/ISI scenario, the probabilities of rupture and leakage are below the acceptance criteria of 1.0x10-6 after 80 years of operation by three orders of magnitude. Furthermore, when the probabilities of rupture and leakage for the alternative inspection schedule are compared to the present ASME Code,Section XI inspection schedule, there is no difference. This indicates that there is no change in risk from the current ASME Code,Section XI schedule to that of the alternative inspection schedule.

4.0 CONCLUSION

S From the PFM and DFM evaluations, the following conclusions are made:

The DFM evaluation demonstrated that a very long operating period (greater than 200 years) is necessary for a postulated initial flaw (with a depth equal to ASME Code,Section XI acceptance standards) to reach the allowable fracture toughness or propagate through 80% of the wall thickness (assumed as leakage for this study). This indicates that all in-scope components at ONS 1/2/3 are very flaw tolerant.

For the ONS 1/2/3 specific inspection history, the PFM evaluation showed that the probabilities of rupture and leakage are significantly below the acceptance criterion of 1.0x10-6 failures per year after 80 years of operation.

File No.: 2100561.303 Revision: 2 Page 11 of 38 F0306-01R4 In the PFM evaluations, conservative number of cycles were used (300 cycles analyzed versus projected actual cycles of 134 cycles for 60 years, or 400 cycles analyzed versus projected actual cycles of 179 for 80 years). In the DFM evaluation, the results indicate that the plant can operate safely for 208 years (the equivalent of 464 cycles). This compares with the actual projected cycles of 134 for 60 years and 179 for 80 years. Hence there are sufficient margins to accommodate any deviations from the projected cycles.

Sensitivity studies involving stress and fracture toughness indicated that when all stresses are increased by a factor of 1.4 or the fracture toughness reduced from 106 ksiin to 72 ksiin, the acceptance criteria are met for both rupture and leakage.

A sensitivity study involving increasing all the stresses by 10% and reducing the fracture toughness from 106 ksiin to 80 ksiin also showed that the probability of rupture and leakage are below the acceptance criteria.

For Item Nos. B2.11 and B2.12, coverage is greater than 90% for all welds and therefore essentially 100% coverage is achieved. For Item No. B3.110, the minimum coverage is 25.2%.

An evaluation using this coverage results in acceptable probabilities of rupture and leakage for this Item No. Furthermore, when compared to the mandated ASME Code,Section XI inspection schedule, there is no change in risk.

5.0 REFERENCES

1. Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905, ADAMS Accession No. ML21021A271.

2. SI Calculation 2100561.302, Rev. 1, Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head.

3. SI Calculation 2100561.304, Rev. 0, Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Safety Relief Nozzle and Top Head.

4. ASME Boiler and Pressure Vessel Code,Section XI, 2017 Edition.

5. SI Calculation FP-ONS-304P, Oconee SI:FatiguePro 4 Baseline Analysis, Startup through 11/3/2020 (U1), 11/21/2019 (U2) and 4/24/2020 (U3), Revision 0. Only non-Proprietary information from this reference was used.

6. SI Calculation 2100561.301, Rev. 0, Duke Plants Insurge/Outsurge Transients.

7. Simonen, F. A. and Johnson, K. I., Effects of Residual Stresses and Underclad Flaws on the Reliability of Reactor Pressure Vessels, PVP-Vol. 251, Reliability and Risk in Pressure Vessels and Piping, ASME PVP Conference, 1993.

File No.: 2100561.303 Revision: 2 Page 12 of 38 F0306-01R4

8. API Standard 579-1/ASME FFS-1, Fitness-for-Service, Second Edition, June 2016.

9. SI-TIFFANY 3.1, Structural Integrity Associates, September 2018.

10. Novetech Corporation, Evaluation of Flaws in Ferritic Piping, EPRI NP-6045, October,1988.

11. C-S Seok, Effect of Temperature on the Fracture Toughness of A-516 Grade 70 Steel, Korean Society of Mechanical Engineers (KSME) International Journal, Vol. 14, No. 1, pp 11 - 18, 2000.

12. V. Mehta, Evaluation of the Fracture Parameters for SA-516 Grade 70 Material, IOSR Journal of Mechanical and Civil Engineering (IOSR-JMCE), e-ISSN: 2278-1684, p-ISSN: 2320-334X, Volume 13, Issue 3 Ver. III (May-Jun. 2016), PP 38-45.

13. V. Mehta, Experimental Study of the Fracture Parameters for the SMAW Joints of SA 516 -

Grade 70 Material, International Journal of Research and Scientific Innovation (IJRSI), Volume IV, Issue VI, June 2017, ISSN 2321-2705.

14. A. Zahoor, Materials and Fracture Mechanics Assessments of Railroad Tank Cars, NISTIR 6266, U. S. Dept. of Commerce, Technology Administration, Materials Performance Group, Metallurgical Division, National Institute of Standards and Technology, Gaithersburg, MD 20899, September 1998.

15. J. G. Early, Metallurgical Analysis of ASTM A212-B Steel Tank Car Head Plate, Report No.

FRA/ORC 81/32, PB 81-205098, National Bureau of Standards, Washington DC 20545, April 1981.

16. R. Roberts and C. Newton, Interpretive Report on Small Scale Test Correlations with KIc Data, Welding Research Council Bulletin 265, February 1981.

17. Structural Integrity Associates Report DEV1806.402, PROMISE 2.0 Theory and Users Manual, Revision 1.

18. Duke Examination Results.xlsx, SI File No. 2100561.207.

File No.: 2100561.303 Revision: 2 Page 13 of 38 F0306-01R4 Table 1: Summary of DFM Design Inputs Input Value Geometry From Section 4 Initial Crack Size 5.2% of the thickness, c/a = 1 Fracture toughness 106 ksiin Fatigue crack growth law ASME Code,Section XI Appendix A, Paragraph A-4300 Operating Transient Stresses From Reference [2]

Operating Cycles From Tables 2 and 3 Residual stresses Cosine curve with 8 ksi peak (Figure 16)

Table 2: ONS 1/2/3 Plant Specific Transient Cycles Used in the DM and PFM Evaluations Transient ONS1/2/3 60-Year Projection Heatup /

Cooldown 132/134/104(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 62/33/41(2)

Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 13, 14 and 15 of [5] scaled down from 80 to 60 years. 134 cycles conservatively used in the DFM evaluations.

2. Loss of Load = Rx Trip No Loss of Flow from Tables 13, 14 and 15 of [5] scaled down from 80 to 60 years. 62 cycles conservatively used in the DFM evaluations.

File No.: 2100561.303 Revision: 2 Page 14 of 38 F0306-01R4 Table 3: Summary of ONS 1/2/3 Plant Specific Insurge/Outsurge Temperature Differences and Numbers of Cycles Used in the DFM Evaluations(1)

T (oF)(2) 60-Year No. of Cycles for Evaluation 400 0

350 0

300 720 250 360 200 1440 Note:

(1) From Reference [6]

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

File No.: 2100561.303 Revision: 2 Page 15 of 38 F0306-01R4 Table 4: Results of the DFM Evaluation Item No.

Component Description Case Identification(1)

Years to Reach KIc of 106 ksi SFprimary = 2.0 SFsecondary = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 925 PRSNV-BW-1C 1783 PRSNV-BW-2A 980 PRSNV-BW-2C 1975 PRSNV-BW-3A 1601 PRSNV-BW-3C 2037 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shall welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 671 PRSHC-BW-4C 208 PRSHC-BW-5A 1648 PRSHC-BW-5C 288 PRSHC-BW-6A 1817 PRSHC-BW-6C 396 PRSHC-BW-7A 1633 PRSHC-BW-7C 402 PRSHC-BW-8A 290 PRSHC-BW-8C 468 PRSHC-BW-9A 232 PRSHC-BW-9C 478 PRSHC-BW-10A 231 PRSHC-BW-10C 899 PRSHC-BW-11A 334 PRSHC-BW-11C 1200 PRSHC-BW-12A 208 PRSHC-BW-12C 336 PRSHC-BW-13A 240 PRSHC-BW-13C 249 Note 1:

The Case Identification terminology is as follows: PR for Pressurizer; SNV for surge nozzle-to-vessel and SHC for head or shell-to-head; BW for B&W design; P1 through P13 represent the crack paths (see Figure 2); C for circumferential part-through-wall crack; and A for axial part-through-wall crack.

File No.: 2100561.303 Revision: 2 Page 16 of 38 F0306-01R4 Table 5: PFM Inputs for ONS 1/2/3 Inspection Scenario No. of Realizations Epistemic = 1, Aleatory = 10 million No. of cracks per weld 1, constant Crack depth distribution PVRUF Crack length distribution NUREG/CR-6817-R1 Fracture toughness (ksiin)

Normal (106,5)

Inspection coverage 100%

PSI Yes ISI 10, 20, 30, 40 and 70 years POD Curve BWRVIP-108, Figure 8-6 Fatigue crack growth law and threshold A-4300, log-normal, Second Parameter = 0.467 Operating Transient Stresses and Cycles From Reference [2]

Uncertainties on transients None Weld residual stresses (ksi)

Cosine Curve (8, 8), constant (not random)

File No.: 2100561.303 Revision: 2 Page 17 of 38 F0306-01R4 Table 6: Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for Oconee Units 1, 2 and 3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70)

Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 106 ksi Stress Multiplier = 1.0 Flaw Density =1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 1.25E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 1.25E-09 1.25E-09 PRSHC-BW-4C 2.50E-09 1.25E-09 PRSHC-BW-5A 1.25E-09 1.25E-09 PRSHC-BW-5C 1.25E-09 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 PRSHC-BW-6C 1.25E-09 1.25E-09 PRSHC-BW-7A 1.25E-09 1.25E-09 PRSHC-BW-7C 1.25E-09 1.25E-09 PRSHC-BW-8A 1.25E-09 1.25E-09 PRSHC-BW-8C 1.25E-09 1.25E-09 PRSHC-BW-9A 1.25E-09 1.25E-09 PRSHC-BW-9C 1.25E-09 1.25E-09 PRSHC-BW-10A 1.25E-09 1.25E-09 PRSHC-BW-10C 1.25E-09 1.25E-09 PRSHC-BW-11A 1.25E-09 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 PRSHC-BW-12A 2.50E-09 1.25E-09 PRSHC-BW-12C 1.25E-09 1.25E-09 PRSHC-BW-13A 1.25E-09 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 18 of 38 F0306-01R4 Table 7: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 106 ksiin Stress Multiplier = 1.0 Flaw Density = 1.0 KIc = 72 ksiin Stress Multiplier = 1.0 Flaw Density = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 1.25E-09 1.25E-09 6.25E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 1.25E-09 1.25E-09 1.00E-08 1.25E-09 PRSHC-BW-4C 2.50E-09 1.25E-09 6.85E-07 1.25E-09 PRSHC-BW-5A 1.25E-09 1.25E-09 5.00E-09 1.25E-09 PRSHC-BW-5C 1.25E-09 1.25E-09 1.23E-07 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-6C 1.25E-09 1.25E-09 2.63E-08 1.25E-09 PRSHC-BW-7A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-7C 1.25E-09 1.25E-09 2.25E-08 1.25E-09 PRSHC-BW-8A 1.25E-09 1.25E-09 1.13E-08 1.25E-09 PRSHC-BW-8C 1.25E-09 1.25E-09 5.00E-09 1.25E-09 PRSHC-BW-9A 1.25E-09 1.25E-09 2.00E-08 1.25E-09 PRSHC-BW-9C 1.25E-09 1.25E-09 6.25E-09 1.25E-09 PRSHC-BW-10A 1.25E-09 1.25E-09 1.63E-08 1.25E-09 PRSHC-BW-10C 1.25E-09 1.25E-09 5.00E-09 1.25E-09 PRSHC-BW-11A 1.25E-09 1.25E-09 1.00E-08 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-12A 2.50E-09 1.25E-09 2.65E-07 1.25E-09 PRSHC-BW-12C 1.25E-09 1.25E-09 2.25E-08 1.25E-09 PRSHC-BW-13A 1.25E-09 1.25E-09 1.41E-07 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 19 of 38 F0306-01R4 Table 8: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Stress Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 106 ksiin Stress Multiplier = 1.0 Flaw Density = 1.0 KIc = 106 ksiin Stress Multiplier = 1.4 Flaw Density = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSHC-BW-4C 2.50E-09 1.25E-09 8.56E-07 1.25E-09 PRSHC-BW-5A 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSHC-BW-5C 1.25E-09 1.25E-09 5.00E-08 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-6C 1.25E-09 1.25E-09 1.13E-08 1.25E-09 PRSHC-BW-7A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-7C 1.25E-09 1.25E-09 1.13E-08 1.25E-09 PRSHC-BW-8A 1.25E-09 1.25E-09 7.50E-09 1.25E-09 PRSHC-BW-8C 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSHC-BW-9A 1.25E-09 1.25E-09 2.63E-08 1.25E-09 PRSHC-BW-9C 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSHC-BW-10A 1.25E-09 1.25E-09 2.13E-08 1.25E-09 PRSHC-BW-10C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-11A 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-12A 2.50E-09 1.25E-09 4.90E-07 1.25E-09 PRSHC-BW-12C 1.25E-09 1.25E-09 2.50E-08 1.25E-09 PRSHC-BW-13A 1.25E-09 1.25E-09 1.36E-07 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 20 of 38 F0306-01R4 Table 9: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness and Stress Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 80 ksi Stress Multiplier = 1.1 Flaw Density = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 5.00E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 8.75E-09 1.25E-09 PRSHC-BW-4C 6.53E-07 1.25E-09 PRSHC-BW-5A 2.50E-09 1.25E-09 PRSHC-BW-5C 8.88E-08 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 PRSHC-BW-6C 2.63E-08 1.25E-09 PRSHC-BW-7A 2.50E-09 1.25E-09 PRSHC-BW-7C 1.88E-08 1.25E-09 PRSHC-BW-8A 7.50E-09 1.25E-09 PRSHC-BW-8C 6.25E-09 1.25E-09 PRSHC-BW-9A 2.00E-08 1.25E-09 PRSHC-BW-9C 3.75E-09 1.25E-09 PRSHC-BW-10A 1.88E-08 1.25E-09 PRSHC-BW-10C 2.50E-09 1.25E-09 PRSHC-BW-11A 6.25E-09 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 PRSHC-BW-12A 3.05E-07 1.25E-09 PRSHC-BW-12C 2.63E-08 1.25E-09 PRSHC-BW-13A 1.08E-07 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 21 of 38 F0306-01R4 Table 10: Sensitivity Study for ISI Examination Coverage for ONS 1/2/3 B3.110 Welds Stress Path ID ASME Section XI Inspection Interval PSI/ISI Scenario: 0,10,20,30,40,50,60,70 Flaw density = 1.0 KIC = 106 ksiin, SD = 5 ksiin Stress multiplier = 1.0 Coverage = 25.2%

Alternate Inspection Interval PSI/ISI Scenario: 0,10,20,30,40,70 Flaw density = 1.0 KIC = 106 ksiin, SD = 5 ksiin Stress multiplier = 1.0 Coverage = 25.2%

Probability of Rupture Probability of Leakage Probability of Rupture Probability of Leakage PRSNV-BW-1A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 22 of 38 F0306-01R4 Figure 1. Modeled Dimensions (From Reference [2])

File No.: 2100561.303 Revision: 2 Page 23 of 38 F0306-01R4 Figure 2: Path Locations (From Reference [2])

File No.: 2100561.303 Revision: 2 Page 24 of 38 F0306-01R4 Figure 3 Through-Wall Stress Distribution for Path P1 [2]

File No.: 2100561.303 Revision: 2 Page 25 of 38 F0306-01R4 Figure 4. Through-Wall Stress Distribution for Path P2 [2]

File No.: 2100561.303 Revision: 2 Page 26 of 38 F0306-01R4 Figure 5. Through-Wall Stress Distribution for Path P3 [2]

File No.: 2100561.303 Revision: 2 Page 27 of 38 F0306-01R4 Figure 6. Through-Wall Stress Distribution for Path P4 [2]

File No.: 2100561.303 Revision: 2 Page 28 of 38 F0306-01R4 Figure 7. Through-Wall Stress Distribution for Path P5 [2]

File No.: 2100561.303 Revision: 2 Page 29 of 38 F0306-01R4 Figure 8. Through-Wall Stress Distribution for Path P6 [2]

File No.: 2100561.303 Revision: 2 Page 30 of 38 F0306-01R4 Figure 9. Through-Wall Stress Distribution for Path P7 [2]

File No.: 2100561.303 Revision: 2 Page 31 of 38 F0306-01R4 Figure 10. Through-Wall Stress Distribution for Path P8 [2]

File No.: 2100561.303 Revision: 2 Page 32 of 38 F0306-01R4 Figure 11. Through-Wall Stress Distribution for Path P9 [2]

File No.: 2100561.303 Revision: 2 Page 33 of 38 F0306-01R4 Figure 12. Through-Wall Stress Distribution for Path P10 [2]

File No.: 2100561.303 Revision: 2 Page 34 of 38 F0306-01R4 Figure 13. Through-Wall Stress Distribution for Path P11 [2]

File No.: 2100561.303 Revision: 2 Page 35 of 38 F0306-01R4 Figure 14. Through-Wall Stress Distribution for Path P12 [2]

File No.: 2100561.303 Revision: 2 Page 36 of 38 F0306-01R4 Figure 15. Through-Wall Stress Distribution for Path P13 [2]

File No.: 2100561.303 Revision: 2 Page 37 of 38 F0306-01R4 Figure 16. Weld Residual Stress Distribution Figure 17. Semi-Elliptical Axial Crack in a Cylinder Model Figure 18. Semi-Elliptical Circumferential Crack in a Cylinder Model

-10

-5 0

5 10 0

0.2 0.4 0.6 0.8 1

x/t Stress (ksi) p

File No.: 2100561.303 Revision: 2 Page 38 of 38 F0306-01R4 Figure 19. The Effect of Temperature on the Fracture Toughness, JIc, of SA-516 Grade 70 Steel [11]

File No.: 2100561.303 Revision: 2 Page A-1 of A-2 F0306-01R4 COMPUTER FILES LISTING

File No.: 2100561.303 Revision: 2 Page A-2 of A-2 F0306-01R4 Zip File Name Description Table6.zip 26 input and 26 output files for the results in Table 6. The filenames are the same as Case Identification.

Table7.zip 26 input and 26 output files for the results in Table 7. The filenames are the same as Case Identification.

Table8.zip 26 input and 26 output files for the results in Table 8. The filenames are the same as Case Identification.

Table9.zip 26 input and 26 output files for the results in Table 9. The filenames are the same as Case Identification.

Table10A.zip 12 input and 12 output files for the results in Table 10 for PSI+10+20+30+40+50+60+70. The filenames are the same as Case Identification.

Table10B.zip 12 input and 12 output files for the results in Table 10 for PSI+10+20+30+40+70. The filenames are the same as Case Identification.

Table4.zip 26 input and 26 output files for the results in Table 4. The filenames are the same as Case Identification.