ND-18-0309, 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) | |
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Michael J.Yox 7825 River Road SOUtnGrn NUCl03r Regulatory Affairs Director Waynesboro, GA 30830 Vogtle3&4 706-848-6459 tel Mak 1 2 2018 Docket Nos.:
52-025 52-026 410-474-8587 cell myox@southernco.com 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 Unit4 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 planfor completing this ITAAC. Southern Nuclear Operating Company will, at a laterdate, 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 Guidelinefor 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 allSection 52.99(c)(1) ITAAC Closure Notifications will be submitted to NRC to support the Commission finding that all acceptance criteria are met priorto 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 identifiedin 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 ASMECode 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 identifiedin Table 2.2.4-1 as ASME Code Section III retain their pressure boundary integrity at their design pressure.
4.b) The piping identified inTable 2.2.4-2 as ASME Code Section III retains its pressure boundary integrity at its design pressure.
5.b) Each ofthe linesidentified in Table 2.2.4-2forwhich functional capability is required is designed to withstand combined normal and seismicdesign 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 ofthe protection from the dynamic effectsofa 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 accordancewith the ASME Code Section III.
Ahydrostatic test will be performed onthe components and piping required bythe ASME Code Section III to be hydrostatically tested.
Inspection will be performed for the existence ofa report verifying thatthe as-built piping meets the requirements for functional capability.
Inspection will be performed for the existence ofan LBB evaluation report oran evaluation report on the protection from effectsofa pipebreak. Section 3.3, Nuclear Island Buildings, contains the designdescriptions 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 LBBevaluation 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 inthe 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 inaccordance 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 inaccordance with the VEGP Updated Final SafetyAnalysis Report (UFSAR) and the ASME Code Section III requirements. The ASME Code Section III certified Design Reports for these components exist and documentthat the as-built components conform to the approved design details. The ASME Section III Design Report foreach component isdocumentedinthe component'scompleted 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 DataReport(s) for the applicable piping system (Reference 1).
The as-builtpiping listedinTable 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 Codeprovisions. Design reconciliation ofthe as-built systems, including installed components, validates that construction completion, including field changesand anynonconforming condition dispositions, isconsistent with and bounded bythe approved design. All applicable fabrication, installation and testing records, as well as, those forthe 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 ultrasonictesting, as required byASME Code Section III) ofthe components' pressure boundary welds are documented inthe Non-destructive Examination Report(s), which support completion ofthe 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 ofthe requiredexamination and testing ofthe item, which includes non-destructive examinations of pressure boundary welds. Satisfactory completion ofthe non-destructive examination of pressure boundarywelds ensures thatthe 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 inaccordance with Reference2 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 SymbolN-Stampand the corresponding piping system Code N-5 Data Report Form(s) (Reference1), 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 supportcompletion ofthe respective Code Stamping and Code N-5 Data Report(s). The completion ofstamping the respectivepiping 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) andthe ASME Code Section III andthatthe 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 ofthe 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 inthe HydrostaticTesting 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 correspondingASME Code Data Reports (certified bythe Authorized Nuclear Inspector)ensures that the components are constructed inaccordance with the Design Specifications and the ASME Code Section III and that the satisfactory completion ofthe hydrostatic pressure testing ofeach component identified in Table2.2.4-1 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 inTable 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.
Ahydrostatic test is performed in accordance with procedureXYZ (as applicable) that complies with the ASME Code (1998 Edition, 2000 Addenda),Section III requirementsto demonstrate that the ASME Code Section III piping identified in Table 2.2.4-2 retains its pressure boundary integrity at its design pressure.
Ahydrostatic test verifies that there are no leaks at welds or piping, and that the pressure boundary integrity is retained at itsdesign 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 Table2.2.4-2 respectively conform with the requirements ofthe 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 pipinglines identified in ITAAC Table 2.2.4-2 that requires functional capabilityis designed to withstand combined normal and seismic design basis loads without a loss of itsfunctional capability. "Functional capability," inthis context, refers to the capabilityof the piping to withstand the effects of earthquakes, without a loss ofsafety function (to conveyfluids from one location to another). Specificfunctional capability requirementsare defined inthe VEGP UFSAR Table 3.9-11 (Reference3).
Piping functional capability is nota specific ASME Code requirement but it is a requirement in the VEGP UFSAR (Reference3). Assuch, information demonstrating that UFSAR functional capability requirements are metis included in the ASME Section ill As-Built Design Reports for safety class piping prepared in accordance with ASME Section III NCA-3550 under theASME 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 thatthe as-built piping systemsare analyzed for functional capability and for compliance with the design specification and ASME Codeprovisions. Design reconciliation ofthe as-built systems validates that construction completion, including field changes and any nonconforming condition dispositions, isconsistent with and bounded by theapproved design. As required by ASME Code, the As-Built Design Report includes the results ofphysical 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 arecomplete and conclude that each of the as-built SGS piping lines for which functional capability isrequired meets therequirements 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 reportexists and concludes that the LBB acceotance criteria are met bv the as-builtSGS 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 toconfirm 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 Bcannotmeetthe LBB acceptance criteria, a pipebreakevaluation 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 BreakHazards 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 reportexists and concludes that the ASME Code Section III requirements are met for non-destructive examination of pressure boundary welds; A reportexists and concludes that the results ofthe hydrostatic test ofthe 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; Areport existsand concludes that each ofthe as-built lines identified in Table2.2.4-2 for which functional capability is required meets the requirementsforfunctional 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, ora pipe breakevaluation report exists and concludes that protectionfrom the dynamiceffects ofa 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 ofall ITAAC findings and associated corrective actions.
This review, which included nowconsolidated 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 Number220]
- 6. Completion Package for Unit 4 ITAAC 2.2.04.02a [COL Index Number220]
U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 9 of 12 Attachment A SYSTEM: Steam Generator System (SGS)
Equipment Name
- Tag No.
- ASME Code Section ill*
ASME III as-built Design Report N-5 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 Steam Generator 01 SGS-PL-V027A Yes XXX N-5 Code Data Report Power-operated Relief Valve Block Motor-operated Valve Steam Generator 02 SGS-PL-V027B Yes XXX N-5 Code Data Report Steam Line Condensate Drain Isolation Valve SGS-PL-V036A Yes XXX N-5 Code Data Report Steam Line Condensate Drain Isolation Valve SGS-PL-V036B Yes XXX N-5 Code Data Report 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 Control Valve SGS-PL-V086A Yes XXX N-5 Code Data Report Steam Line Condensate Drain Control Valve SGS-PL-V086B Yes XXX N-5 Code Data Report 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)
Equipment Name
- Tag No.
- ASME Code Section ill*
ASME III as-built Design Report N-5 Report Main Feedwater Isolation Valve SGS-PL-V057S Yes XXX N-5 Code Data Report Startup Feedwater Isolation Motor-operated Valve SGS-PL-V067A Yes XXX N-5 Code Data Report Startup Feedwater Isolation Motor-operated Valve SGS-PL-V067S Yes XXX N-5 Code Data Report Steam Generator Slowdown Isolation Valve SGS-PL-V074A Yes XXX N-5 Code Data Report Steam Generator Slowdown Isolation Valve SGS-PL-V074S Yes XXX N-5 Code Data Report Steam Generator Slowdown Isolation Valve SGS-PL-V075A Yes XXX N-5 Code Data Report Steam Generator Slowdown Isolation Valve SGS-PL-V075S Yes XXX N-5 Code Data Report Power-operated Relief Valve SGS-PL-V233A Yes XXX N-5 Code Data Report Power-operated ReliefValve SGS-PL-V233S Yes XXX N-5 Code Data Report Main Steam Isolation Valve Svpass Isolation SGS-PL-V240A Yes XXX N-5 Code Data Report Main Steam Isolation Valve Svoass Isolation SGS-PL-V240S Yes XXX N-5 Code Data Report 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 Thermal Relief Valve SGS-PL-V257A Yes XXX N-5 Code Data Report Main Feedwater Thermal Relief Valve SGS-PL-V257S Yes XXX N-5 Code Data Report Startup Feedwater Thermal Relief Valve SGS-PL-V258A Yes XXX N-5 Code Data Report Startup Feedwater Thermal Relief Valve SGS-PL-V258S Yes XXX N-5 Code Data Report "Excerpts from COLAppendixC 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)
Line Name*
Line Number*
ASME Code Section III*
Leak Before Break*
Functional Capability Required*
ASME III As-Built Design Report LBB evaiua^tion/
pipe break evaluation N-5 Report Main Feedwater Line SGS-PL-L002A, L002B Yes No No XXX N/A N-5 Code Data Report Main Feedwater Line SGS-PL-L003A, L003B Yes No No XXX N/A N-5 Code Data Report Startup Feedwater Line SGS-PL-L004A, L004B Yes No No XXX N/A N-5 Code Data Report Startup Feedwater line SGS-PL-L005A, L005B Yes No No XXX N/A N-5 Code Data Report Main Steam Line (within containment)
SGS-PL-L006A, L006B Yes Yes Yes XXX YYY N-5 Code Data Report Main Steam Line (outside of containment)
SGS-PL-L006A, L006B Yes No Yes XXX N/A N-5 Code Data Report Main Steam Line SGS-PL-L007A, L007B Yes No No XXX N/A N-5 Code Data Report Safety Valve Inlet Line SGS-PL-L015A, L015B, L015C, L015D, L015E, L015F, L015G, L015H, L015J, L015K, L015L, L015M Yes No Yes XXX N/A N-5 Code Data Report Safety Valve Discharge Line SGS-PL-L018A, L018B, L018C, L018D, L018E, L018F, L018G, L018H, L018J, L018K, L018U L018M Yes No Yes XXX N/A N-5 Code Data Report Power-operated Relief Block Valve Inlet Line SGS-PL-L024A, L024B Yes No No XXX N/A N-5 Code Data Report
U.S. Nuclear Regulatory Commission ND-18-0309 Enclosure Page 12 of 12 Attachment B SYSTEM: Steam Generator System (SGS)
Line Name*
Line Number*
ASME Code Section ill*
Leak Before Break*
Functional Capability Required*
ASME III As-Built Design Report LBB evaluation/
pipe break evaluation N-5 Report Power-operated Relief Valve Inlet Line SGS-PL-L014A, L014S Yes No No XXX N/A N-5 Code Data Report Main Steam Isolation Valve Bypass Inlet Line SGS-PL-L022A, L022S Yes No No XXX N/A N-5 Code Data Report Main Steam Isolation Valve Bypass Outlet Line SGS-PL-L023A, L023S Yes No No XXX N/A N-5 Code Data Report Main Steam Condensate Drain Line SGS-PL-L021A, L021S Yes No No XXX N/A N-5 Code Data Report Steam Generator Slowdown Line SGS-PL-L009A, L009S Yes No No XXX N/A N-5 Code Data Report Steam Generator Slowdown Line SGS-PL-L027A, L027S Yes No No XXX N/A N-5 Code Data Report Steam Generator Slowdown Line SGS-PL-L010A, L010S Yes No No XXX N/A N-5 Code Data Report
'Excerpts from COL Appendix C, Table 2.2.4-2