W3F1-2024-0008, Proposed Alternative WF3-RR-24-01 for Examinations of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds

Proposed Alternative WF3-RR-24-01 for Examinations of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds

ML24078A374
Person / Time
Site:	Waterford
Issue date:	03/18/2024
From:	Couture P Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
W3F1-2024-0008
Download: ML24078A374 (1)

Text

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

W3F1-2024-0008 10 CFR 50.55a

March 18, 2024

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

Subject:

Proposed Alternative WF3-RR-24-01 for Examinations of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds

Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38

In accordance with 10 CFR 50.55a(z)(1), Entergy Operations, Inc. (Entergy) hereby requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative for Waterford Steam Electric Station, Unit 3 (WF3). Specifically, the proposed alternative concerns Class 1, Examination Category B-B, Pressure-Retaining Welds in Vessels Other Than Reactor Vessels,

Item Number B2.11 - Pressurizer, Shell-to-Head Welds, Circumferential, and B2.12 -

Pressurizer, Shell-to-Head Welds, Longitudinal, and Class 1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Item Number B3.110 - Pressurizer, nozzle-to-vessel welds.

The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer at WF3 from the current American Society of Mechanical Engineers (ASME) Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is currently scheduled to end on December 18, 2044. This equates to an extension of 27 years, 18 days from the end of the third inservice inspection interval (November 30, 2017) at which time all ASME Code,Section XI, Division 1 requirements were satisfied. Entergy requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that the alternative provides an acceptable level of quality and safety.

The Enclosure to this letter provides the Alternative Request WF3-RR-24-01 with Enclosure, Attachment 1 providing the WF3 Plant-Specific Applicability and Enclosure, Attachment 2 providing the Results of Industry Survey.

Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213 W3F1-2024-0008 Page 2 of 2

Entergy requests approval of the proposed Alternative Request by February 28, 2025, to support the WF3 Refueling Outage (RF26). The proposed changes would be implemented during RF26.

This letter contains no new regulatory commitments.

Should you have any questions or require additional information, please contact me at 601-368-5102.

Respectfully,

Phil Couture

PC/chm

Enclosure:

Alternative Request WF3-RR-24-01

Attachments to

Enclosure:

1. Plant-Specific Applicability for WF3

2. Results of Industry Survey

cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - WF3 NRC Project Manager - WF3

Enclosure

W3F1-2024-0008

Alternative Request WF3-RR-24-01

W3F1-2024-0008 Enclosure Page 1 of 16

TABLE OF CONTENTS

1.0 ASME CODE COMPONENTS AFFECTED...................................................................... 2 2.0 APPLICABLE CODE EDITION AND ADDENDA.............................................................. 3 3.0 APPLICABLE CODE REQUIREMENT............................................................................. 3 4.0 REASON FOR REQUEST................................................................................................ 3 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE....................................................... 3 5.1 Technical Basis.............................................................................................................. 4 5.1.1 Applicability of the Degradation Mechanism Evaluation in Reference [9.1] to WF3 4 5.1.2 Applicability of the Stress Analysis in References [9.1] to WF3.............................. 5 5.1.3 Applicability of the Flaw Tolerance Evaluation in References [9.1] to WF3............ 6 5.1.4 Inspection History.................................................................................................... 9 5.1.5 Industry Survey....................................................................................................... 9 5.1.6 Performance Monitoring.......................................................................................... 9 5.1.7 Conclusion............................................................................................................... 9 6.0 DURATION OF PROPOSED ALTERNATIVE................................................................. 10 7.0 PRECEDENTS................................................................................................................ 10 7.1 Salem Units 1 and 2..................................................................................................... 10 7.2 Other Approved Actions Related to Inspections of PZR Welds and Components....... 11 7.3 Other Studies by the Industry....................................................................................... 12 8.0 ACRONYMS................................................................................................................... 12

9.0 REFERENCES

................................................................................................................ 13 10.0 ATTACHMENTS............................................................................................................. 16

W3F1-2024-0008 Enclosure Page 2 of 16

ALTERNATIVE REQUEST WF3-RR-24-01

1.0 ASME CODE COMPONENTS AFFECTED American Society of Mechanical Engineers (ASME)

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

Waterford Unit 3 (WF3)

ASME ASME Item No. Component ID Component Description Category B-B B2.11 05-002 Bottom head to lower shell circ. weld B-B B2.11 05-008 Top head to shell weld B-B B2.12 05-003 Lower shell long. weld 90-degrees(1)

B-B B2.12 05-004 Lower shell long. weld 180-degrees(1)

B-B B2-12 05-006 Upper shell long. weld 90-degrees(1)

B-B B2-12 05-007 Upper shell long. weld 180-degrees(1)

B-D B3.110 05-009 Surge nozzle to bottom head weld B-D B3.110 05-010 Spray nozzle to top head weld B-D B3.110 05-011 Safety nozzle to top head weld 45-degrees

B-D B3.110 05-012 Safety nozzle to top head weld 135-degrees

B-D B3.110 05-013 Safety nozzle to top head weld 225-degrees

Note 1: The applicable portion of the longitudinal seam weld is where it intersects the associated Item No. B2.11 (shell to head) weld.

W3F1-2024-0008 Enclosure Page 3 of 16

2.0 APPLICABLE CODE EDITION AND ADDENDA The fourth 10-year inservice inspection (ISI) interval Code of record for WF3 is the 2007 Edition of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI through the 2008 Addenda, "Rules for Inservice Inspection of Nuclear Power Plant Components."

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

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

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

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

4.0 REASON FOR REQUEST

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

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

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

report concluded that the current ASME Code,Section XI ISI examinations can be deferred for some time with no impact to plant safety. Based on the conclusions of the EPRI report supplemented by plant-specific evaluations contained herein, Entergy is requesting an ISI examination deferral for the subject welds. The Reference [9.1] report was developed consistent with the recommendations provided in EPRIs White Paper on suggested content for PFM submittals [9.12] and NRC Regulatory Guide 1.245 for PFM submittals and associated technical basis [9.13, 9.14].

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE

For WF3, Entergy is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

W3F1-2024-0008 Enclosure Page 4 of 16

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

WF3 still has its original pressurizer. The pressurizer welds and components received the required PSI examinations prior to service followed by ISI examinations through the first period of the current fourth inspection interval.

The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer at WF3 from the current ASME Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is currently scheduled to end on December 18, 2044. This equates to an extension of 27 years, 18 days from the end of the third inservice inspection interval (November 30, 2017) at which time all ASME Code,Section XI, Division 1 requirements were satisfied.

5.1 Technical Basis A summary of the key aspects of the technical basis for this request is summarized below. The applicability of the technical basis to WF3 is shown in Attachment 1.

5.1.1 Applicability of the Degradation Mechanism Evaluation in Reference [9.1] to WF3 An evaluation of degradation mechanisms that could potentially impact the reliability of the pressurizer welds was performed in Reference [9.1]. The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the pressurizer welds covered in this request. This observation was a cknowledged by the NRC in Section 2, page 3, second paragraph of the Reference [9.16] Safety Evaluation (SE) for Salem Units 1 & 2.

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

As part of the technical basis in Reference [9.1], a comprehensive industry survey involving 74 PWR units was conducted to determine the degradation history of these components. The survey reviewed examination results from the start of plant operation.

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

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

W3F1-2024-0008 Enclosure Page 5 of 16

5.1.2 Applicability of the Stress Analysis in References [9.1] to WF3 Finite element analysis (FEA) was performed in Reference [9.1] to determine the stresses in the WF3 pressurizer welds covered in this request. The analysis was performed using representative Westinghouse plant geometries (which bounds CE plants), bounding transients, and typical materi al properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to WF3 is demonstrated in Attachment 1 and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [9.1] stress analysis are compared to those of WF3 in Tables 1 and 2:

Table 1 WF3 Pressurizer Shell Dimensions Plant Shell ID Shell/Clad Shell Ri/t (in) Thk (in)

EPRI Report (Table 4-4 of [9.1]) 84(1)(2) 3.75/0.063(1)(2) 11.2 WF3 96.25(3) 4.875/0.22(3) 9.87 Notes:

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

2. Value from Table 4-4 of Reference [9.1] (based on OD) converted to ID-based value for comparison with WF3.

3. Reference [9.20].

Table 2 WF3 Pressurizer Nozzle Dimensions Plant Surge Nzl Surge Nzl Surge Nzl SRV Nzl ID SRV Nzl SRV Nzl Ri/t ID (in) Thk (in) Ri/t (in) Thk (in)

EPRI Report (Table 4-5 of [9.1]) 12.44(1) 3.27(1) 1.9 5.625(2) 1.19(2) 2.363 WF3 12(3) 3.875(3) 1.55 5.19(4) 1.41(4) 1.84

Notes:

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

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

3. Reference [9.21].

4. PDF page 376 (of 788) of Reference [9.22].

As noted by the NRC in Section 5.1, page 7, fourth paragraph of the Salem Safety Evaluation (SE) [9.16], the dominant stress is the pressure stress. Therefore, the variation in the Ri/t ratio determined in Tables 1 and 2 can be used to scale up the stresses of the Reference [9.1] report to obtain the plant-specific stresses for each unit and component. Since all pressurizer welds at WF3 listed in Section 1 are shell welds, W3F1-2024-0008 Enclosure Page 6 of 16

Table 1 applies. From this table, the stress ratio (R i/t) of WF3 relative to the that used in the EPRI report is (9.87/11.2) = 0.88. Therefore, the stresses at WF3 are less than those considered in the EPRI report, indicating that the Westinghouse plant geometry used in the stress analysis of Reference [9.1] bounds the CE plants.

In the selection of the transients in Section 5 of Reference [9.1] and the subsequent stress analyses in Section 7, test conditions beyond a system leakage test were not considered since pressure tests at WF3 are performed at normal operating conditions.

No hydrostatic testing had been performed at WF3 since the unit went into operation.

5.1.3 Applicability of the Flaw Tolerance Evaluation in References [9.1] to WF3 Flaw tolerance evaluations were performed in Reference [9.1] consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. Since the Westinghouse pressurizer configuration considered in Reference [9.1] is bounding relative to the CE pressurizer design, the results of the flaw tolerance evaluation can be conservatively applied to the WF3 pressurizer. The results of the PFM analyses indicate that, after a preservice inspection (PSI) followed by subsequent in-service inspections (ISI), the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year is met.

The PFM analysis in Reference [9.1] was performed using the PRobabilistic OptiMization of I nSp Ection ( PROMISE) Version 1.0 software, developed by Structural Integrity Associates. As part of the NRCs review of Southern Nuclears alternative request, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan dated May 14, 2020 (ML20128J311). The PFM analysis in Reference [9.1] was performed using the PROMISE Version 2.0 software which has not been audited by the NRC. The only technical difference between the two versions is that in PROMISE Version 1.0, the user-specified examination coverage is applied to all inspections, whereas in PROMISE Version 2.0, the examination coverage can be specified by the user uniquely for each inspection. In both Versions 1.0 and 2.0, the software assumes 100% coverage for the PSI examination. The NRC staff found the use PROMISE Version 2.0 acceptable in Section 3.1, page 5, fourth paragraph of the Reference [9.16] SE for Salem. Note that the assumption below of a 30-year ISI deferral is conservative compared to the end of currently licensed operating life for WF3.

For the WF3 pressurizer, PSI examinations have been performed followed by ISI examinations for three complete ISI intervals. The PSI/ISI scenario for the WF3 pressurizer is therefore (PSI+10+20+30+60). This scenario was not specifically considered in the Reference [9.1] PFM evaluations in combination with key variables, as evaluated by the NRC in Section 4.0 (page 6) of the Reference [9.16] Safety Evaluation.

Therefore, a new PFM evaluation was performed for this limiting PSI/ISI scenario using PROMISE Version 2.0, the same version used for the evaluations in the EPRI report

[9.1]. The evaluations were performed for the critical Case ID from Reference [9.1]

(PRSHC-BW-2C) with a combination of the most dominant parameters (stress and fracture toughness) as identified by the NRC in Section 4.0 (page 6) and Section 10 (page 19) of Reference [9.14]. Since all welds under consideration are shell welds, a flaw density of 1.0 was used in the evaluation. This flaw density value was found acceptable by the NRC in Section 9.6 of Reference [9.16]. A fracture toughness of 200 ksiin with a standard deviation of 5 ksi in was used, as recommended by the NRC in Section 10 (page 19) of Reference [9.16]. A stress multiplier of 2.1 was used in the W3F1-2024-0008 Enclosure Page 7 of 16

evaluation. This stress multiplier was conservatively chosen such that probability of rupture or leakage will be close to the acceptance criteria of 1.0x10 -6 after 80 years. As discussed above, a conservative stress multiplier of 0.88 can be applied to the WF3 pressurizer components and therefore the stress multiplier of 2.1 used in the evaluation is very conservative. The results of the evaluation are presented in Table 3.

Table 3 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Weld Flaw Density for 80 Years for the WF3 Pressurizer Welds (Case ID PRSHC-BW-2C from Reference [9.1])

Probability per Year for Combined Case KIC = 200 ksiin.,

Time SD = 5 ksiin.

(year) Stress Multiplier = 2.1 Nozzle Flaw Density = 1 PSI+10+20+30+60 Rupture Leak 10 3.90E07 1.00E08 20 3.55E07 5.00E09

30 2.40E07 3.33E09 40 1.80E07 2.50E09

50 1.44E07 2.00E09 60 1.23E07 1.67E09

70 1.06E07 1.43E09 80 9.25E08 1.25E09

The plant-specific PFM evaluation presented above for WF3 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 multiplier applied in Table 3 is greater than the WF3 R i/t ratio shown in Table 1 relative to that of the model in the EPRI report and therefore the analysis in Table 3 is conservative. It should be noted that the evaluation incorporates conservative assumptions with regard to the PSI/ISI scenarios. Furthermore, the evaluation was performed for 80 years, which is longer than the deferral being sought by Entergy in this Request for Alternative.

In the PFM evaluations in Reference [9.1], the Pressure Vessel Research Facility Users Facility (PVRUF) initial flaw size distribution was used. This distribution is applicable to thick vessels and not to relatively thin vessels like pressurizers. This issue was raised by the NRC in RAI No. 4 in Reference [9.17]. In response to this RAI, various initial flaw size distributions were used in a sensitivity study [9.16] which showed that regardless of which distribution was used, the conclusions of Reference [9.1] remain the same. This W3F1-2024-0008 Enclosure Page 8 of 16

was found acceptable by the NRC in Section 9.1, page 15, last paragraph of the SE for Salem [9.16].

An evaluation was performed to show acceptability of the low KIC values at the beginning and ending of the heatup/cooldown transient to address the NRC RAI No. 4 for PSEG in Reference [9.17], using the maximum RT NDT value of 60oF allowed by BTP 5-3 [9.19].

The RTNDT value of 60oF bounds the limiting RTNDT value of 30oF in Attachment 1 for the pressurizer materials at WF3. The evaluation was performed for the most critical Case ID (PRSHC-BW-2C) from Reference [9.1], similar to that performed in Reference [9.18]

to respond to RAI 1 in Reference [9.17]. The three flaw sizes evaluated in the PSEG RAI response are also considered. The results are shown Figure 1. As seen in this figure, the calculated applied stress intensity factors are bounded with margin by the corresponding KIC calculated as a function of temperature (based on an RT NDT of 60oF) throughout the transient, for all three flaw depths. The heatup/cooldown transient also includes the leak test; hence, this addresses the adequacy of the temperature of the leak test to ensure that the applied stress intensity factor is below the fracture toughness during the leak test.

Figure 1.

Applied K vs. Fracture Toughness as a Function of Temperature for Case ID PRSHC-BW-2C (RTNDT = 60oF)

The DFM evaluation in Table 8-4 of Reference [9.1] provides verification of the above PFM results for WF3 by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to ASME Code,Section XI acceptance W3F1-2024-0008 Enclosure Page 9 of 16

standards to grow to a depth where the maximum stress intensity factor (K) exceeds the ASME Code,Section XI allowable fracture toughness.

5.1.4 Inspection History As described in Section 8.3.4.1 of Reference [9.1], preservice inspection (PSI) refers to the superset of the 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 WF3 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 1. As shown in the attachment, the minimum examination coverage for the WF3 components is 48.3%, which is greater than the 37% determined to be acceptable per Section 10 of the Salem Safety Evaluation (Reference [9.16]). As shown in Attachment 1, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

5.1.5 Industry Survey The inspection history for these components as obtained from an industry survey is presented in Attachment 2. The results of the survey indicate that these components are very flaw tolerant.

5.1.6 Performance Monitoring To provide additional defense in depth to the Request for Alternative, Entergy will adopt a performance monitoring plan during the requested deferral period. The components listed below have received a PSI examination and at least one ISI examination during the third 10-year interval. Since a 27-year, 18-day deferral is being requested for these component examinations, Entergy would like to propose performing the examinations (previously planned for the fourth 10-year interval) prior to end of currently licensed operating life, which is scheduled to end on December 18, 2044.

Components: 05-007,05-008 5.1.7 Conclusion It is concluded that the pressurizer pressure-retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis report [9.1], supplemented by plant-specific evaluations performed as part of this Request for Alternative, demonstrate that using conservative PSI/ISI inspection scenarios for all plants, the NRC safety goal of 1.0x10 -6 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to WF3 is demonstrated in Attachment 1. 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. Attachment 1 shows the examination history for the pressurizer welds examined in the two most recent 10-year inspection intervals.

W3F1-2024-0008 Enclosure Page 10 of 16

In addition to the required PSI examinations for these pressurizer welds, WF3 has performed multiple ISI examinations through the current 10-year inspection interval at each plant.

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

Examination coverage for WF3 was greater than 48%, which was determined to be acceptable per Section 10 of the Salem Safety Evaluation (Reference [9.16]). 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.

Finally, as discussed in Reference [9.3], for situations where no active degradation mechanism is present, it was concluded that subsequent ISI examinations do not provide additional value after PSI has been performed and the inspection volumes have been confirmed to be free of defects.

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

6.0 DURATION OF PROPOSED ALTERNATIVE The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer at WF3 from the current ASME Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is currently scheduled to end on December 18, 2044. This equates to an extension of 27 years, 18 days from the end of the third inservice inspection interval (November 30, 2017) at which time all ASME Code,Section XI, Division 1 requirements were satisfied.

7.0 PRECEDENTS 7.1 Salem Units 1 and 2 The following previous submittal has been made by PSEG Nuclear to provide relief from the ASME Code,Section XI, Examination Category B-B (Item Nos. B2.11 and B2.12) and Category B-D (Item No. B3.110) surface and volumetric examinations based on the Reference 9.1 technical basis report:

PSEG Nuclear Letter to NRC - "Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12,"

(ML20218A587), dated August 5, 2020, [9.15].

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

NRC Letter to PSEG Nuclear - "Salem Generating Station Unit Nos. 1 and 2 -

Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200,"

(ML21145A189), dated June 10, 2021, [9.16].

W3F1-2024-0008 Enclosure Page 11 of 16

7.2 Other Approved Actions Related to Inspections of PZR Welds and Components In addition, follows is a list of other Relief Requests and other precedents related to inspections of pressurizer welds and components:

NRC Letter to Entergy - "Indian Point Nuclear Generating Unit No. 2 - Relief Request No. RR-01," (ML072130487), dated September 5, 2007, [9.23].

NRC Letter to Entergy - "Indian Point Nuclear Generating Unit No. 2 - Safety Evaluation for Relief Request No. IP2-ISI-RR-01," (ML16179A178), dated September 14, 2016, [9.24].

NRC Letter to 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," (ML110691154), dated April 26, 2011,

[9.25].

NRC Letter to Southern - "Second Ten-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant Units 1 and 2," (ML011640178), dated June 20, 2001, [9.26].

NRC Letter to Exelon - "Braidwood Station Units 1 and 2 - Relief from Requirements of the ASME Code for the Third 10-Year Interval of Inservice Inspection," (ML13016A515), dated January 30, 2013, [9.27].

NRC Letter to Duke - "Catawba Nuclear Station, Unit 1 - Request for Relief 05-CN-004, Limited Weld Examinations During End-of-Cycle 15 Refueling Outage,"

(ML062390020), dated September 25, 2006, [9.28].

NRC Letter to Duke - "Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 11-CN-001 for the Third 10-Year Inservice Inspection Interval,"

(ML12228A723), dated August 20, 2012, [9.29].

NRC Letter to Duke - "Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 14-CN-001, American Society of Mechanical Engineers (ASME)Section XI Volumetric Examination Requirements," (ML14295A532), dated October 30, 2014, [9.30].

Duke Letter to NRC - "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," (ML11279A035),

dated September 21, 2011, [9.31].

Dominion Letter to NRC - "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,"

(ML101130187), dated April 19, 2010, [9.32].

Progress Letter 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," (ML090540055), dated February 5, 2009,

[9.33].

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Progress Letter 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," (ML092740063), dated September 24, 2009, [9.34].

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

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

Based on work performed in BWRVIP-108 [9.6] and BWRVIP-241 [9.8], the NRC approved the reduction of BWR vessel nozzle-to-shell weld examinations (Item No. B3.90 for BWRs from 100% to a 25% sample of each nozzle type every 10 years) in References [9.7] and [9.9]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702

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

8.0 ACRONYMS ASME American Society of Mechanical Engineers B&W Babcock and Wilcox BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Program CE Combustion Engineering CFR Code of Federal Regulations DFM Deterministic fracture mechanics EAF Environmentally assisted fatigue EPRI Electric Power Research Institute FAC Flow accelerated corrosion FEA Finite element analysis FW Feedwater ISI Inservice Inspection MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system W3F1-2024-0008 Enclosure Page 13 of 16

O.D. Outside diameter POD Probability of detection PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor PZR Pressurizer SCC Stress corrosion cracking WEC Westinghouse Electric Company

9.0 REFERENCES

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

9.2 Duke Letter to NRC, "Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)," (ML23048A148), dated February 17, 2023 9.3 ASME - 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., dated 1992 and 1998.

9.4 WCAP - WCAP-16168-NP-A, Rev. 3, "Risk-Informed extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011.

9.5 NRC Letter to PWROG - "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," (ML111600303),

dated July 26, 2011 9.6 EPRI - BWRVIP-108, "BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii," EPRI, Palo Alto, CA 2002, 1003557.

9.7 NRC Letter to EPRI - 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)," (ML073600374), dated December 19, 2007 9.8 EPRI - 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.

W3F1-2024-0008 Enclosure Page 14 of 16

9.9 NRC Letter to EPRI - 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),"

(ML13071A240 and ML13071A233), dated April 19, 2013 9.10 ASME - Code Case N-702, "Alternate Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," ASME Code Section XI, Division 1, Approval Date February 20, 2004.

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

9.12 EPRI Letter to NRC - BWR Vessel & Internals Project (BWRVIP) Memo No.

2019-016, "White Paper on Suggested Content for PFM Submittals to the NRC,"

(ML19241A545), dated February 27, 2019.

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

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

9.15 PSEG Nuclear Letter to NRC - "Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12,"

(ML20218A587), dated August 5, 2020.

9.16 NRC Letter to PSEG Nuclear - "Salem Generating Station Unit Nos. 1 and 2 -

Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200,"

(ML21145A189), dated June 10, 2021.

9.17 NRC Letter to PSEG Nuclear - "Requests for Additional Information Regarding Salem Generating Station Units Nos. 1 and 2 Regarding Alternative for Examination of ASME Section XI, Category B-B, Item Number B2.11 and B2.12,"

(ML21043A144), dated February 11, 2021.

9.18 PSEG Nuclear Letter to NRC - "Response to Request for Additional Information for Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12," (ML21102A024), dated April 12, 2021.

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

9.20 CE - E-74370-671-002. "General Arrangement Waterford III 96" I.D. Pressurizer,"

Combustion Engineering Inc. dated 1972.

9.21 Entergy - UT Vessel Examination Report No. 2-ISI-UT-14-017 (file 2R23 05-009 UT.pdf), (Note: ANO2 replacement pressurizer dimensions also applicable to WF3 pressurizer), dated May 27, 2014.

9.22 Entergy - Engineering Report No. ER-ANO-2002-0836-003, "ANO-2 Pressurizer Replacement Project," Revision 0 (Note: ANO-2 replacement pressurizer dimensions also applicable to WF3 pressurizer).

9.23 NRC letter to Entergy - "Indian Point Nuclear Generating Unit No. 2 - Relief Request No. RR-01," (ML072130487), dated September 5, 2007.

W3F1-2024-0008 Enclosure Page 15 of 16

9.24 NRC Letter to Entergy - "Indian Point Nuclear Generating Unit No. 2 - Safety Evaluation for Relief Request No. IP2-ISI-RR-01," (ML16179A178), dated September 14, 2016.

9.25 NRC Letter to 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," (ML110691154), dated April 26, 2011.

9.26 NRC Letter to Southern - "Vogtle Electric Generating Plant Units 1 and 2 -

Second Ten-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33," (ML011640178), dated June 20, 2001.

9.27 NRC Letter to Exelon - "Braidwood Station Units 1 and 2 - Relief from Requirements of the ASME Code for the Third 10-Year Interval of Inservice Inspection," (ML13016A515), dated January 30, 2013.

9.28 NRC Letter to Duke - "Catawba Nuclear Station, Unit 1 - Request for Relief 05-CN-004, Limited Weld Examinations During End-of-Cycle 15 Refueling Outage," (ML062390020), dated September 25, 2006.

9.29 NRC Letter to Duke - "Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 11-CN-001 for the Third 10-Year Inservice Inspection Interval,"

(ML12228A723), dated August 20, 2012.

9.30 NRC Letter to Duke - "Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 14-CN-001, American Society of Mechanical Engineers (ASME)

Section XI Volumetric Examination Requirements," (ML14295A532), dated October 30, 2014.

9.31 Duke Letter to NRC - "McGuire Nuclear Station Units 1 and 2, Relief Request Serial # 11-MN-001, Limited Weld Examinations for Refueling Outage 1EOC20 and 2EOC19," (ML11279A035), dated September 21, 2011.

9.32 Dominion Letter to NRC - "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," (ML101130187), dated April 19, 2010.

9.33 Progress Letter 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," (ML090540055), dated February 5, 2009.

9.34 Progress Letter 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," (ML092740063), dated September 24, 2009.

W3F1-2024-0008 Enclosure Page 16 of 16

10.0 ATTACHMENTS

1. Plant-Specific Applicability for WF3

2. Results of Industry Survey

Enclosure, Attachment 1

W3F1-2024-0008

Plant-Specific Applicability for WF3

W3F1-2024-0008 Enclosure, Attachment 1 Page 1 of 8

PLANT-SPECIFIC APPLICABILITY FOR WF3

Section 9 of Reference [1-1] provides requirements that must be demonstrated to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for WF3 is provided in Table 1-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report are applicable to WF3.

Table 1-1 Applicability of Reference [1-1] Representative Analyses to WF3 Pressurizer Surge Nozzle and Bottom Head Welds (Item Nos. B2.11 and B3.110)

Category Requirement from Reference [1-1] Applicability to WF3 General The plant-specific pressurizer general transients As shown in Table 1-3, the WF3 Requirements and cycles must be bounded by those shown in general transients are bounded by the Table 5-6 for a 60-year operating life. It should transients listed in Table 5-6 of be noted that the number of cycles were Reference [1-1].

extrapolated to 80 years in the evaluations.

The materials of the pressurizer surge nozzle, The WF3 pressurizer shell and surge bottom head and shell must be low alloy ferritic nozzle are fabricated from SA-533 steels which conform to the requirements of Grade B Class 1 and SA-508-2 ASME Code,Section XI, Appendix G, material, respectively (per Reference Paragraph G-2110. [1-3]). The RTNDT value for this material is -30°F except for the bottom head material at 30°F. All material RTNDT values are bounded by the value used in the EPRI report.

The SA-533 Grade B Class 1 and SA-508-2 material are low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific pressurizer surge nozzle, The WF3 pressurizer surge nozzle and Requirements bottom head and shell weld configurations must bottom head weld configurations are conform to those shown in Figure 1-1 (Item No. shown in Figures 1-1, 1-3 and 1-4 and B2.11) and Figures 1-3 and 1-4 (Item No. show conformance with the Figures B3.110) of Reference [1-1]. shown in Reference [1-1].

The plant-specific dimensions of the pressurizer As shown in Table 1-2, the WF3 surge nozzle and shell must be within the range pressurizer shell and surge nozzle of values listed in Table 9-1 of Reference [1-1]. dimensions are within the range of values listed in Table 9-1 of Reference

[1-1].

W3F1-2024-0008 Enclosure, Attachment 1 Page 2 of 8

Category Requirement from Reference [1-1] Applicability to WF3 The plant-specific Insurge/Outsurge transient As shown in Table 1-4, the WF3 definitions (temperature difference between the Insurge/Outsurge transients are pressurizer shell and the pressurizer surge bounded by the transients listed in nozzle fluid temperature and associated number Table 5-10 of Reference [1-1].

of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference [1-1].

Pressurizer Top Head Welds (Item Nos. B2.11 and B3.110)

Category Requirement from Reference [1-1] Applicability to WF3 General The plant-specific pressurizer general transients As shown in Table 1-3, the WF3 Requirements and cycles must be bounded by those shown in general transients are bounded by the Table 5-6 of Reference [1-1] for a 60-year transients listed in Table 5-6 of operating life. It should be noted that the Reference [1-1].

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

The materials of the pressurizer top head The WF3 pressurizer shell and surge nozzles, head and shell must be low alloy ferritic nozzle are fabricated from SA-533 steels which conform to the requirements of Grade B Class 1 and SA-508-2 ASME Code,Section XI, Appendix G, materials, respectively (per Reference Paragraph G-2110. [1-3]). The RTNDT value for this material is -30°F except for the top head material at 30°F. All material RTNDT values are bounded by the value used in the EPRI report.

The SA-533 Grade B Class 1 and SA-508-2 materials are low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific pressurizer top head nozzles, The WF3 pressurizer spray nozzle, Requirements head and shell weld configurations must safety valve nozzle and top head weld conform to those shown in Figure 1-1 (Item No. configurations are shown in Figures B2.11) and Figures 1-3 and 1-4 (Item No. 1-1, 1-3, 1-5 and 1-6 and show B3.110) of Reference [1-1]. conformance with the Figures shown in Reference [1-1].

The plant-specific dimensions of the pressurizer As shown in Table 1-2, the WF3 top head nozzles and shell must be within the pressurizer shell and top head nozzle range of values listed in Table 9-1 of Reference dimensions are within the range of

[1-1]. values listed in Table 9-1 of Reference

[1-1].

W3F1-2024-0008 Enclosure, Attachment 1 Page 3 of 8

Table 1-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with WF3 Component Geometric Parameter For a CE Plant (From Table WF3 9-1 of Reference [1-1]) Dimensions Pressurizer Shell Inside Diameter (in) Must be between 90 and 102 96.25 NPS of piping or component (e.g.,

Surge Nozzle reducer) attached to nozzle safe-end Must be between 10 and 14 12 (in) (1)

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

Nozzle reducer) attached to nozzle safe-end Must be between 4 and 6 6 (in) (1)

NPS of piping or component (e.g.,

Spray Nozzle reducer) attached to nozzle safe-end Must be between 4 and 6 4 (in) (1)

Note:

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

Table 1-3 Comparison of WF3 General Transients to Requirements in Reference [1-1]

Number of Cycles for WF3 Transient 60 Years from Table 5-6 60-Year Projection of Reference [1-1]

Heatup / Cooldown 300 144/144(1)

Loss of Load (Large Step Load Decrease, Loss of Power, 360 187(1)(2)

Loss of Flow, Reactor Trip)

Notes:

1. Table 4.3-1 (page 4.3-4) of Reference [1-2]

2. Loss of Load = Reactor Trip

W3F1-2024-0008 Enclosure, Attachment 1 Page 4 of 8

Table 1-4 Comparison of WF3 Insurge/Outsurge Transients to Requirements in Reference [1-1]

60-Year No. of Cycles WF3 T ( oF)(1) From Table 5-10 of Reference [1-1] (For 60-Year Westinghouse and CE Plants) Projection

330 600 11(2)

320 3,000 1,119(2)

103 1,500 1,130(2)

Notes:

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

2. Reference [1-7].

Table 1-5 WF3 Inspection History Item Component Exam Interval/Period/Outage Exam Coverage Relief No. ID Date Results(1) Request B2.11 05-002 11/04/2009 Third / 1st / RF16 NRI 95.75% na B2.11 05-002 1/12/2019 Fourth / 1st / RF22 NRI 99.6% na B2.11 05-008 10/28/2012 Third / 2nd / RF18 NRI >90% na B2.12 05-003 11/04/2009 Third / 1st / RF16 NRI 100% na B2.12 05-003 01/12/2019 Fourth / 1st / RF22 NRI 100% na B2.12 05-007 10/28/2012 Third / 2nd / RF18 NRI >90% na B3.110 05-009 04/16/2011 Third / 2nd / RF17 NRI 56.2% na B3.110 05-009 01/13/2019 Fourth / 1st / RF22 NRI 51.2% na B3.110 05-010 10/27/2009 Third / 1st / RF16 AI 80.4% na B3.110 05-010 01/13/2019 Fourth / 1st / RF22 NRI 48.3% na B3.110 05-011 10/27/2009 Third / 1st / RF16 AI 80.7% na B3.110 05-011 01/11/2019 Fourth / 1st / RF22 AI 55.6% na B3.110 05-012 10/27/2009 Third / 1st / RF16 AI 80.7% na B3.110 05-012 01/11/2019 Fourth / 1st / RF22 AI 55.6% na B3.110 05-013 10/27/2009 Third / 1st / RF16 NRI 80.7% na B3.110 05-013 01/11/2019 Fourth / 1st / RF22 NRI 55.6% na Notes:

1. NRI = no recordable indications; AI = indications identified but determined to be acceptable.

W3F1-2024-0008 Enclosure, Attachment 1 Page 5 of 8

Figure 1-1 WF3 Pressurizer Vessel [1-4]

W3F1-2024-0008 Enclosure, Attachment 1 Page 6 of 8

Figure 1-2 WF3 Upper Shell-to-Head Weld (B2.11) From Reference [1-5])

Figure 1-3 WF3 Lower Shell-to-Head Weld (B2.11) from Reference [1-5])

W3F1-2024-0008 Enclosure, Attachment 1 Page 7 of 8

Figure 1-4 WF3 Pressurizer Surge Nozzle from Reference [1-6])

Figure 1-5 WF3 Pressurizer Spray Nozzle from Reference [1-6])

W3F1-2024-0008 Enclosure, Attachment 1 Page 8 of 8

Figure 1-6 WF3 Pressurizer Safety Valve Nozzle from Reference [1-6])

References

1-1. EPRI - Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019, 3002015905.

1-2. Entergy - "Waterford Steam Electric Station, Unit 3, License Renewal Application,"

(ML16088A331 thru A335), dated March 23, 2016.

1-3. Entergy - WSES-FSAR-UNIT-3, Table 5.2-8. "Waterford 3 Pressurizer Materials Fracture Toughness Data."

1-4. Entergy - SEP-ISI-104, Revision 14. "Program Section for ASME Section XI, Division 1 WF3 Inservice Inspection Program," Entergy Nuclear Engineering Programs, dated July 19, 2023.

1-5. Combustion Engineering Inc. (CE) - E-74370-661-001. "As-Built Dimensions Waterford III 96" I.D. Pressurizer." Combustion Engineering Inc, 1972.

1-6. CE - E-74370-671-002. "General Arrangement Waterford III 96" I.D. Pressurizer."

Combustion Engineering Inc, 1972.

1-7. Structural Integrity Engineering, Inc. (SI) - SI Calculation No. 2200654.302, "Waterford Unit 3 (WSES) Plant Loading/Unloading and Pressurizer Insurge/Outsurge Transients,"

Revision 0.

Enclosure, Attachment 2

W3F1-2024-0008

Results of Industry Survey

W3F1-2024-0008 Enclosure, Attachment 2 Page 1 of 1

RESULTS OF INDUSTRY SURVEY

Overall Industry Inspection Summary The results of an industry survey of past inspections of pressurizer welds are summarized in Reference [2-1]. Table 2-1 provides a summary of the combined survey results for Item Nos. B2.11, B2.12, B2.21, B2.22 and B3.110. The results identify that pressurizer examination of the items adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international PWR units responded to the survey and provided information representing all PWR plant designs currently in operation in the U.S. This included 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR nuclear steam supply system (NSSS) vendors (i.e., Babcock and Wilcox (B&W), Combustion Engineering (CE), and Westinghouse). A total of 1,162 examinations for the components of the affected Item Nos. were conducted on PWR pressurizer components.

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

Table 2-1 Summary of Survey Results

Item No. No. of Examinations No. of Reportable Indications

B2.11 269 4 (1)

B2.12 269 0 B2.21 4 0 B2.22 30 0 B3.110 590 0 Note:

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

References 2-1. EPRI - Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019, 3002015905.