ND-18-0309, Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.2.04.02a (Index Number 220)

Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.2.04.02a (Index Number 220)

ML18072A155
Person / Time
Site:	Vogtle
Issue date:	03/12/2018
From:	Yox M Southern Nuclear Operating Co
To:	Document Control Desk, Office of New Reactors
References
ITAAC 2.2.04.02a, ND-18-0309
Download: ML18072A155 (15)

Text

kI I Michael J. Yox 7825 River Road SOUtnGrn NUCl03r Regulatory Affairs Director Waynesboro, GA 30830 Vogtle3&4 706-848-6459 tel 410-474-8587 cell myox@southernco.com Mak 1 2 2018 Docket Nos.: 52-025 52-026 ND-18-0309 10CFR 52.99(c)(3)

U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Unit 3 and Unit 4 Notice of Uncompleted ITAAC 225-davs Prior to Initial Fuel Load Item 2.2.04.02a [Index Number 220]

Ladies and Gentlemen; Pursuant to 10 CFR 52.99(c)(3), Southern Nuclear Operating Company hereby notifies the NRC that as of March 5, 2018, Vogtle Electric Generating Plant (VEGP) Unit 3 and Unit 4 Uncompleted Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Item 2.2.04.02a [Index Number 220]

has not been completed greater than 225-days priorto initial fuel load. The Enclosure describes the plan for completing this ITAAC. Southern Nuclear Operating Company will, at a later date, provide additional notifications for ITAAC that have not been completed 225-days prior to initial fuel load.

This notification is informed by the guidance described in NEI 08-01, Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52, which was endorsed by the NRC in Regulatory Guide 1.215. In accordance with NEI 08-01, this notification includes ITAAC for which required inspections, tests, or analyses have not been performed or have been only partially completed. All ITAAC will be fully completed and all Section 52.99(c)(1) ITAAC Closure Notifications will be submitted to NRC to support the Commission finding that all acceptance criteria are met prior to plant operation, as required by 10 CFR 52.103(g).

This letter contains no new NRC regulatory commitments.

If there are any questions, please contact Tom Petrak at 706-848-1575.

Respectfully submitted, I

Michael J. Yox Regulatory Affairs Director Vogtle 3 &4

Enclosure:

Vogtle Electric Generating Plant (VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC 2.2.04.02a [Index Number 220]

MJY/KJD/amw

U.S. Nuclear Regulatory Commission ND-18-0309 Page 2 of 3 To:

Southern Nuclear Operating Company I Georgia Power Company Mr. D. A. Bost (w/o enclosures)

Mr. M. D. Rauckhorst (w/o enclosures)

Mr. M. D. Meier Mr. D. H. Jones (w/o enclosures)

Mr. D. L. McKinney Mr. M. J. Yox Mr. D. L. Fulton Mr. J. B. Klecha Mr. F. H. Willis Ms. A. L. Pugh Mr. A. 8. Parton Mr. W. A. Sparkman Mr. 0. E. Morrow Ms. K. M. Stacy Mr. M. K. Washington Mr. J. P. Redd Ms. A. 0. Chamberlain Mr. D. R. Culver Mr. T. G. Petrak Document Services RTYPE: VND.LI.LOO File AR.01.02.06 cc:

Nuclear Regulatory Commission Mr. W. Jones (w/o enclosures)

Ms. J. M. Heisserer Mr. C. P. Patel Mr. M. E. Ernstes Mr. G. J. Khouri Mr. T. E. Chandler Ms. S. E. Temple Ms. P. Braxton Mr. N. D. Karlovich Mr. A. J. Lerch Mr. C. J. Even Mr. F. D. Brown Mr. B. J. Kemker Ms. A. E. Rivera-Varona Ms. L. A. Kent Mr. P. B. Donnelly Oolethorpe Power Corporation Mr. R. B. Brinkman Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. S. M. Jackson

U.S. Nuclear Regulatory Commission ND-18-0309 Page 3 of 3 Dalton Utilities Mr. T. Bundros Westinahouse Electric Company. LLC Dr. L. Oriani (w/o enclosures)

Mr. D. 0. Durham (w/o enclosures)

Mr. M. M. CorlettI Ms. L. G. Iller Mr. D. Hawkins Ms. J. Monahan Mr. J. L. Coward Ms. N. E. Deangelis Other Mr. J. E. Hosier, Bechtel Power Corporation Ms. L. Matis, Tetra Tech NUS, Inc.

Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc.

Mr. 8. Roetger, Georgia Public Service Commission Ms. 8. W. Kernizan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. 8. Blanton, Baich Bingham

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 1 of 12 Southern Nuclear Operating Company ND-18-0309 Enclosure Vogtle Electric Generating Plant (VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC 2.2.04.02a [Index Number 220]

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 2 of 12 ITAAC Statement Design Commitment:

2.a) The components identified in Table 2.2.4-1 as ASME Code Section III are designed and constructed in accordance with ASME Code Section III requirements.

2.b) The piping identified in Table 2.2.4-2 as ASME Code Section III is designed and constructed in accordance with ASME Code Section III requirements.

3.a) Pressure boundary welds in components identified in Table 2.2.4-1 as ASME Code Section III meet ASME Code Section III requirements.

3.b) Pressure boundary welds in piping identified in Table 2.2.4-2 as ASME Code Section III meet ASME Code Section III requirements.

4.a) The components identified in Table 2.2.4-1 as ASME Code Section III retain their pressure boundary integrity at their design pressure.

4.b) The piping identified in Table 2.2.4-2 as ASME Code Section III retains its pressure boundary integrity at its design pressure.

5.b) Each of the lines identified in Table 2.2.4-2for which functional capability is required is designed to withstand combined normal and seismic design basis loads without a loss of its functional capability.

6. Each of the as-built lines identified in Table 2.2.4-2 as designed for LBB meets the LBB criteria, or an evaluation is performed of the protection from the dynamic effects of a rupture of the line.

Inspections. Tests. Analyses:

Inspection will be conducted ofthe as-built components and piping as documented in the ASME design reports.

Inspection ofthe as-built pressure boundary welds will be performed in accordance with the ASME Code Section III.

Ahydrostatic test will be performed on the components and piping required bythe ASME Code Section III to be hydrostatically tested.

Inspection will be performed for the existence of a report verifying that the as-built piping meets the requirements for functional capability.

Inspection will be performed for the existence of an LBB evaluation report or an evaluation report on the protection from effects of a pipe break. Section 3.3, Nuclear Island Buildings, contains the design descriptions and inspections, tests, analyses, and acceptance criteria for protection from the dynamic effects of pipe rupture.

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 3 of 12 Acceptance Criteria:

The ASME Code Section III design reports exist for the as-built components and piping identified in Tables 2.2.4-1 and 2.2.4-2 as ASME Code Section III.

A report exists and concludes that the ASME Code Section III requirements are met for non destructive examination of pressure boundary welds.

A report exists and concludes that the results of the hydrostatic test of the components and piping identified in Tables 2.2.4-1 and 2.2.4-2 as ASME Code Section III conform with the requirements of the ASME Code Section III.

A report exists and concludes that each of the as-built lines identified in Table 2.2.4-2 for which functional capability is required meets the requirements for functional capability.

An LBB evaluation report exists and concludes that the LBB acceptance criteria are met by the as-built SGS piping and piping materials, or a pipe break evaluation report exists and concludes that protection from the dynamic effects of a line break is provided.

ITAAC Completion Description This ITAAC requires inspections, tests, and analyses be performed and documented to ensure the Steam Generator System (SGS) components and piping listed in the Combined License (COL) Appendix C, Table 2.2.4-1 (Attachment A) and Table 2.2.4-2 (Attachment B) that are identified as American Society of Mechanical Engineers (ASME) Code Section III, Leak Before Break (LBB), or Functional Capability Required are designed and constructed in accordance with applicable requirements.

2.a and 2.b) The ASME Code Section III design reports exist for the as-built components and Dioina identified in Tables 2.2.4-1 and 2.2.4-2 as ASME Code Section III.

Each component listed in Table 2.2.4-1 as ASME Code Section III is fabricated in accordance with the VEGP Updated Final Safety Analysis Report (UFSAR) and the ASME Code Section III requirements. The ASME Code Section III certified Design Reports for these components exist and document that the as-built components conform to the approved design details. The ASME Section III Design Report for each component is documented inthe component's completed ASME Section III Code Data Report. The individual component ASME Section III Code Data Reports are documented on the ASME Section III N-5 Code Data Report(s) for the applicable piping system (Reference 1).

The as-built piping listed in Table 2.2.4-2 including the components listed in Table 2.2.4-1 as ASME Code Section III, are subjected to a reconciliation process (Reference 2), which verifies that the as-built piping are analyzed for applicable loads (e.g., stress reports) and for compliance with all design specification and Code provisions. Design reconciliation ofthe as-built systems, including installed components, validates that construction completion, including field changes and any nonconforming condition dispositions, is consistent with and bounded bythe approved design. All applicable fabrication, installation and testing records, as well as, those for the related Quality Assurance (OA) verification/inspection activities, which

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 4 of 12 confirm adequate construction in compliance with the ASME Code Section III and design provisions, are referenced in the N-5 data report and/or its sub-tier references.

The applicable ASME Section III N-5 Code Data Report(s), which include the location of the certified Design Reports for all the components listed in Table 2.2.4-1 (Attachment A) and piping listed in Table 2.2.4-2 (Attachment B) as ASME Code Section III, exist and conclude that these installed components are designed and constructed (including their installation within the applicable as-built piping system) in accordance with the ASME Code (1998 Edition, 2000 Addenda and 1989 Edition, 1989 Addenda),Section III requirements as applicable, as described in UFSAR Subsection 5.2.1 (Reference 3). The N-5 Code Data Reports for the piping system(s) containing the components listed in the Table 2.2.4-1 and Table 2.2.4-2 are identified in Attachments A and B, respectively.

3.a and 3.b) A report exists and concludes that the ASME Code Section III reauirements are met for non-destructive examination of pressure boundary welds.

Inspections are performed in accordance with ASME Code Section III (1998 Edition, 2000 Addenda) to demonstrate that as-built pressure boundary welds in components identified in Table 2.2.4-1 as ASME Code Section III meet ASME Code Section III requirements (i.e., no unacceptable indications).

The applicable non-destructive examinations (including liquid penetrant, magnetic particle, radiographic, and ultrasonic testing, as required by ASME Code Section III) of the components' pressure boundary welds are documented in the Non-destructive Examination Report(s), which support completion of the respective ASME Section III N-5 Code Data Report(s) certified by the Authorized Nuclear Inspector, as listed in Attachment A.

Per ASME Code Section III, Subarticle NCA-8300, "Code Symbol Stamps," the N-5 Code Data Report(s) (Reference 1) documents satisfactory completion of the required examination and testing of the item, which includes non-destructive examinations of pressure boundary welds. Satisfactory completion of the non-destructive examination of pressure boundarywelds ensures that the pressure boundary welds in components identified in Table 2.2.4-1 as ASME Code Section III meet ASME Code Section III requirements.

An inspection is performed in accordance with Reference 2 to demonstrate that the as-built pressure boundary welds in piping identified in Table 2.2.4-2 (Attachment B) as ASME Code Section III meet ASME Code Section III requirements (i.e., no unacceptable indications). This portion ofthe ITAAC is completewhen the piping identified in Table 2.2.4-2, which is encompassed within the respective piping system Code Symbol N-Stamp and the corresponding piping system Code N-5 Data Report Form(s) (Reference 1), is complete. The non-destructive examinations (including visual inspection, liquid penetrant, magnetic particle, radiographic, and ultrasonic testing, as required byASME Code Section III) ofthe piping pressure boundary welds are documented in the Non-destructive Examination Report(s) within the piping system's supporting data package, which support completion of the respective Code Stamping and Code N-5 Data Report(s). The completion of stamping the respective piping system along with the corresponding ASME Code N-5 Data Report Form(s) (certified by the Authorized Nuclear Inspector) ensure that the piping is constructed in accordance with the design specification(s) and the ASME Code Section III and that the satisfactory completion of

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 5 of 12 the non-destructive examinations of piping pressure boundary welds for the pipe lines identified in Table 2.2.4-2 meet ASME Code Section III requirements and are documented in the Non-destructive Examination Report(s) within the supporting data packages.

4.a and 4.b1 A report exists and concludes that the results of the hvdrostatic test of the components and piping identified in Tables 2.2.4-1 and 2.2.4-2 as ASME Code Section III conform with the requirements of the ASME Code Section III.

A hydrostatic test is performed by the vendor to demonstrate that the components identified in Table 2.2.4-1 (Attachment A) as ASME Code Section III retain their pressure boundary integrity at their design pressure. The completion of the N-5 Data Reports is governed by Reference 2.

This portion of the ITAAC is complete once each component identified in Table 2.2.4-1 has their individual Code Symbol N-Stamp and corresponding Code Data Report (Reference 1) completed, and the components are installed into the respective Code Symbol N-Stamped piping system and documented on the corresponding N-5 Code Data Report(s) (Reference 1).

The hydrostatic testing results of the component's pressure boundary are documented in the Hydrostatic Testing Report(s) within the supporting component's data package, which support completion of the respective Code Stamping and Code Data Report(s).

The completion of stamping the individual components and the respective piping system along with the corresponding ASME Code Data Reports (certified by the Authorized Nuclear Inspector) ensures that the components are constructed in accordance with the Design Specifications and the ASME Code Section III and that the satisfactory completion of the hydrostatic pressure testing of each component identified in Table 2.2.4-1 as ASME Code Section III is documented in the Hydrostatic Testing Report(s) within the supporting data packages and meets ASME Code Section III requirements.

This ITAAC also verifies that the piping identified in Table 2.2.4-2 (Attachment B)fully meets all applicable ASME Code,Section III requirements and retains its pressure boundary integrity at its design pressure.

A hydrostatic test is performed in accordance with procedureXYZ (as applicable) that complies with the ASME Code (1998 Edition, 2000 Addenda),Section III requirements to demonstrate that the ASME Code Section III piping identified in Table 2.2.4-2 retains its pressure boundary integrity at its design pressure.

A hydrostatic test verifies that there are no leaks at welds or piping, and that the pressure boundary integrity is retained at its design pressure. The hydrostatic testing results ofthe pipe lines are documented in the Hydrostatic Testing Report(s). The Hydrostatic Testing Report(s) supports completion ofthe ASME Section III N-5 Code Data Report(s) forthe applicable piping system (i.e., SGS) (Reference 1).

The applicable ASME Section III N-5 Code Data Report(s) (Reference 1) identified in Attachments A and B documents that the results of the hydrostatic testing of the components and piping identified in Table 2.2.4-1 and Table 2.2.4-2 respectively conform with the requirements of the Code (1998 Edition, 2000 Addenda),Section III.

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 6 of 12 5.b) A report exists and concludes that each of the as-built lines Identified in Table 2.2.4-2 for which functional caoabilitv is reouired meets the reouirements for functional caoabilitv.

An inspection is performed of the ASME Section III as-built piping design report XXX to verify that the report demonstrates that each of the SGS piping lines identified in ITAAC Table 2.2.4-2 that requires functional capability is designed to withstand combined normal and seismic design basis loads without a loss of its functional capability. "Functional capability," in this context, refers to the capabilityof the piping to withstand the effects of earthquakes, without a loss of safety function (to convey fluids from one location to another). Specificfunctional capability requirementsare defined in the VEGP UFSAR Table 3.9-11 (Reference 3).

Piping functional capability is not a specific ASME Code requirement but it is a requirement in the VEGP UFSAR (Reference 3). As such, information demonstrating that UFSAR functional capability requirements are met is included in the ASME Section ill As-Built Design Reports for safety class piping prepared in accordance with ASME Section III NCA-3550 under the ASME Boiler & Pressure Vessel Code (1998 Edition, 2000 Addenda)Section III requirements. The as-built piping systems are subjected to a reconciliation process (Reference 2), which verifies that the as-built piping systems are analyzed for functional capability and for compliance with the design specification and ASME Code provisions. Design reconciliation ofthe as-built systems validates that construction completion, including field changes and any nonconforming condition dispositions, is consistent with and bounded by the approved design. As required by ASME Code, the As-Built Design Report includes the results of physical inspection ofthe piping and reconciliation to the design pipe stress report.

Inspections of the ASME Code Section III As-Built Piping Design Reports (Reference 4) for the SGS piping lines identified in Table 2.2.4-2 are complete and conclude that each of the as-built SGS piping lines for which functional capability is required meets the requirements for functional capability. The ASME Section III As-Built Piping Design Reports for each of the as-built SGS piping lines in Table2.2.4-2 are identified in Attachment B.

6. An LBB evaluation report exists and concludes that the LBB acceotance criteria are met bv the as-built SGS oipina and oioino materials, or a pioe break evaluation reoort exists and concludes that protection from the dvnamic effects of a line break is orovided.

inspections are performed for the as-built lines identified in Table 2.2.4-2 (Attachment B) to verify that each of the as-built lines designed for LBB meets the LBB criteria, or an evaluation is performed of the protection from the dynamic effects of a rupture of the line. VEGP COL Appendix 0, Section 3.3, Nuclear Island Buildings, contains the design descriptions and inspections, tests, analyses, and acceptance criteria for protection from the dynamic effects of pipe rupture.

LBB evaluations are performed as described in UFSAR subsection 3.6.3 to confirm that the as-built SGS piping (and corresponding piping materials) identified in Attachment Ameet the LBB acceptance criteria described in the UFSAR, Appendix 3B, Leak-Before-Break Evaluation of the API 000 Piping (Reference 3). In cases where an as-built SGS piping line in Attachment Bcannot meet the LBB acceptance criteria, a pipe break evaluation is performed which concludes that protection from the dynamic effects of a line break is provided. The pipe break evaluation criteria is discussed in UFSAR, Section 3.6.4.1, Pipe Break Hazards Analysis

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 7 of 12 (Reference 3) and is documented as a pipe rupture hazards analysis report (pipe break evaluation report).

Inspections are performed to verify that LBB as-built piping evaluation reports for the SGS piping (and corresponding piping materials) identified in Attachment B conclude that the as-built piping analysis is bounded by the applicable bounding analysis curves provided in Appendix 3B of the UFSAR (Reference 3). The results are documented in either the applicable ASME Section III as-built piping design report(s) or in separate LBB evaluation report(s). For cases where an as-built SGS piping line in Attachment B cannot meet the LBB acceptance criteria, inspections are performed to verify that a pipe rupture hazards analysis evaluation report (pipe break evaluation report) exists which concludes that protection from the dynamic effects of a line break is provided.

The applicable ASME Section III as-built piping design report(s), LBB evaluation report(s), or pipe rupture hazards analysis report(s) (pipe break evaluation report(s)) exist and are identified in Attachment B.

References 1 and 4 provide the evidence that the ITAAC Acceptance Criteria requirements listed below are met:

The ASME Code Section III design reports exist for the as-built components and piping identified in Tables 2.2.4-1 and 2.2.4-2 as ASME Code Section III;

• A report exists and concludes that the ASME Code Section III requirements are met for non-destructive examination of pressure boundary welds;

• A report exists and concludes that the results ofthe hydrostatic test of the components and piping identified in Tables 2.2.4-1 and 2.2.4-2 as ASME Code Section III conform with the requirements of the ASME Code Section III;

• A report exists and concludes that each ofthe as-built lines identified in Table2.2.4-2 for which functional capability is required meets the requirements for functional capability; and

• An LBB evaluation report exists and concludes that the LBB acceptance criteria are met by the as-built SGS piping and piping materials, or a pipe breakevaluation report exists and concludes that protectionfrom the dynamic effects of a line break is provided.

References 1 and 4 are available for NRC inspection as part of the Unit 3 and Unit 4 ITAAC 2.2.04.02a Completion Packages (References 5 and 6, respectively).

List of ITAAC Findings In accordance with plant procedures for ITAAC completion. Southern NuclearOperating Company (SNC) performed a review of all ITAAC findings and associated corrective actions.

This review, which included now consolidated ITAAC Indexes 221, 222, 223, 224, 225, 229, and

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 8 of 12 230, found there are no relevant ITAAC findings associated with this ITAAC.

References (available for NRG inspection)

1. SGS ASME N-5 Code Data Repoil(s)

2. APP-GW-GAP-139, "Westinghouse/WECTEC ASME N-5 Interface Procedure"

3. VEGP 3&4 Updated Final Safety Analysis Report

a. Subsection 5.2.1 - Compliance with Codes and Code Cases,

b. Table 3.9 Piping Functional Capability - ASME Class 1, 2, and 3,

c. Subsection 3.6.3 - Leak before Break Evaluation Procedures

d. Subsection 3.6.4.1- Pipe Break Hazards Analysis

e. Appendix 3B - Leak-Before-Break Evaluation of the API 000 Piping

4. SGS ASME III As Built Design Report(s)

5. Completion Package for Unit 3 ITAAC 2.2.04.02a [COL Index Number 220]

6. Completion Package for Unit 4 ITAAC 2.2.04.02a [COL Index Number220]

7. NEI 08-01, "Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52"

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 9 of 12 Attachment A SYSTEM: Steam Generator System (SGS)

ASME III as-ASME Code Equipment Name

• Tag No.

• built Design N-5 Report Section ill*

Report Main Steam Safety Valve SG01 SGS-PL-V030A Yes XXX N-5 Code Data Report Main Steam Safety Valve SG02 SGS-PL-V030B Yes XXX N-5 Code Data Report Main Steam Safety Valve SG01 SGS-PL-V031A Yes XXX N-5 Code Data Report Main Steam Safety Valve SG02 SGS-PL-V031B Yes XXX N-5 Code Data Report Main Steam Safety Valve SG01 SGS-PL-V032A Yes XXX N-5 Code Data Report Main Steam Safety Valve SG02 SGS-PL-V032B Yes XXX N-5 Code Data Report Main Steam Safety Valve SG01 SGS-PL-V033A Yes XXX N-5 Code Data Report Main Steam Safety Valve SG02 SGS-PL-V033B Yes XXX N-5 Code Data Report Main Steam Safety Valve SG01 SGS-PL-V034A Yes XXX N-5 Code Data Report Main Steam Safety Valve SG02 SGS-PL-V034B Yes XXX N-5 Code Data Report Main Steam Safety Valve SG01 SGS-PL-V035A Yes XXX N-5 Code Data Report Main Steam Safety Valve SG02 SGS-PL-V035B Yes XXX N-5 Code Data Report Power-operated Relief Valve Block Motor-operated Valve SGS-PL-V027A Yes XXX N-5 Code Data Report Steam Generator 01 Power-operated Relief Valve Block Motor-operated Valve SGS-PL-V027B Yes XXX N-5 Code Data Report Steam Generator 02 Steam Line Condensate Drain XXX N-5 Code Data Report SGS-PL-V036A Yes Isolation Valve Steam Line Condensate Drain SGS-PL-V036B Yes XXX N-5 Code Data Report Isolation Valve Main Steam Line Isolation Valve SGS-PL-V040A Yes XXX N-5 Code Data Report Main Steam Line Isolation Valve SGS-PL-V040B Yes XXX N-5 Code Data Report Steam Line Condensate Drain SGS-PL-V086A Yes XXX N-5 Code Data Report Control Valve Steam Line Condensate Drain SGS-PL-V086B Yes XXX N-5 Code Data Report Control Valve Main Feedwater Isolation Valve SGS-PL-V057A Yes XXX N-5 Code Data Report

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 10 of 12 Attachment A SYSTEM: Steam Generator System (SGS)

ASME III as-ASME Code Equipment Name

• Tag No.

• built Design N-5 Report Section ill*

Report Main Feedwater Isolation Valve SGS-PL-V057S Yes XXX N-5 Code Data Report Startup Feedwater Isolation SGS-PL-V067A Yes XXX N-5 Code Data Report Motor-operated Valve Startup Feedwater Isolation SGS-PL-V067S Yes XXX N-5 Code Data Report Motor-operated Valve Steam Generator Slowdown XXX N-5 Code Data Report SGS-PL-V074A Yes Isolation Valve Steam Generator Slowdown XXX N-5 Code Data Report SGS-PL-V074S Yes Isolation Valve Steam Generator Slowdown SGS-PL-V075A Yes XXX N-5 Code Data Report Isolation Valve Steam Generator Slowdown SGS-PL-V075S Yes XXX N-5 Code Data Report Isolation Valve Power-operated Relief Valve SGS-PL-V233A Yes XXX N-5 Code Data Report Power-operated Relief Valve SGS-PL-V233S Yes XXX N-5 Code Data Report Main Steam Isolation Valve SGS-PL-V240A Yes XXX N-5 Code Data Report Svpass Isolation Main Steam Isolation Valve SGS-PL-V240S Yes XXX N-5 Code Data Report Svoass Isolation Main Feedwater Control Valve SGS-PL-V250A Yes XXX N-5 Code Data Report Main Feedwater Control Valve SGS-PL-V250S Yes XXX N-5 Code Data Report Startup Feedwater Control Valve SGS-PL-V255A Yes XXX N-5 Code Data Report Startup Feedwater Control Valve SGS-PL-V255S Yes XXX N-5 Code Data Report Main Feedwater XXX N-5 Code Data Report SGS-PL-V257A Yes Thermal Relief Valve Main Feedwater SGS-PL-V257S Yes XXX N-5 Code Data Report Thermal Relief Valve Startup Feedwater SGS-PL-V258A Yes XXX N-5 Code Data Report Thermal Relief Valve Startup Feedwater SGS-PL-V258S Yes XXX N-5 Code Data Report Thermal Relief Valve "Excerpts from COL AppendixC Table 2.2.4-1

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 11 of 12 Attachment B SYSTEM: Steam Generator System (SGS)

ASME ASME III LBB Leak Functional Code As-Built evaiua^tion/ N-5 Line Name* Line Number* Before Capability Section Design pipe break Report Break* Required*

III* Report evaluation N-5 Code Main Feedwater SGS-PL-Yes No No XXX N/A Data Line L002A, L002B Report N-5 Code Main Feedwater SGS-PL-Yes No No XXX N/A Data Line L003A, L003B Report N-5 Code Startup SGS-PL-No No XXX N/A Data Yes Feedwater Line L004A, L004B Report N-5 Code Startup SGS-PL-No No XXX N/A Data Yes Feedwater line L005A, L005B Report Main Steam Line N-5 Code SGS-PL- Data (within Yes Yes Yes XXX YYY L006A, L006B Report containment)

Main Steam Line N-5 Code SGS-PL-(outside of Yes No Yes XXX N/A Data L006A, L006B Report containment)

N-5 Code SGS-PL- Data Main Steam Line Yes No No XXX N/A L007A, L007B Report SGS-PL-L015A, L015B, L015C, L015D, N-5 Code Safety Valve L015E, L015F, Yes No Yes XXX N/A Data Inlet Line Report L015G, L015H, L015J, L015K, L015L, L015M SGS-PL-L018A, L018B, L018C, L018D, N-5 Code Safety Valve XXX N/A Data L018E, L018F, Yes No Yes Discharge Line Report L018G, L018H, L018J, L018K, L018U L018M Power-operated N-5 Code SGS-PL- N/A Data Relief Block Yes No No XXX L024A, L024B Report Valve Inlet Line

U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 12 of 12 Attachment B SYSTEM: Steam Generator System (SGS)

ASME ASME III LBB Leak Functional Code As-Built evaluation/ N-5 Line Name* Line Number* Before Capability Section Design pipe break Report Break* Required*

ill* Report evaluation Power-operated N-5 Code SGS-PL-Relief Valve Inlet Yes No No XXX N/A Data L014A, L014S Report Line Main Steam N-5 Code SGS-PL-Isolation Valve Yes No No XXX N/A Data L022A, L022S Report Bypass Inlet Line Main Steam N-5 Code Isolation Valve SGS-PL-Yes No No XXX N/A Data Bypass Outlet L023A, L023S Report Line Main Steam N-5 Code SGS-PL-Condensate Yes No No XXX N/A Data L021A, L021S Report Drain Line N-5 Code Steam Generator SGS-PL-Yes No No XXX N/A Data Slowdown Line L009A, L009S Report N-5 Code Steam Generator SGS-PL-Yes No No XXX N/A Data Slowdown Line L027A, L027S Report N-5 Code Steam Generator SGS-PL-Yes No No XXX N/A Data Slowdown Line L010A, L010S Report

'Excerpts from COL Appendix C, Table 2.2.4-2