ND-19-1348, Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 3.3.00.02a.i.a (Index Number 760)

Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 3.3.00.02a.i.a (Index Number 760)

ML19326C865
Person / Time
Site:	Vogtle
Issue date:	11/22/2019
From:	Yox M Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ND-19-1348
Download: ML19326C865 (13)

Text

Michael J. Yox 7825 River Road

^Southern Nuclear Regulatory Affairs Director Vogtle 3 & 4 Waynesboro, GA 30830 706-848-6459 tel NOV 2 2 2019 Docket Nos.: 52-025 52-026 ND-19-1348 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 3.3.00.02a.i.a (Index Number 7601 Ladies and Gentlemen:

Pursuant to 10 CFR 52.99(c)(3), Southern Nuclear Operating Company hereby notifies the NRC that as of November 15, 2019, Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Uncompleted Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) Item 3.3.00.02a.i.a

[Index Number 760] has not been completed greater than 225-days prior to initial fuel load. The Enclosure describes the plan for completing ITAAC 3.3.00.02a.i.a [Index Number 760]. 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.

Southern Nuclear Operating Company(SNC) previously submitted Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load for Item 3.3.00.02a.i.a [Index Number 760] ND 1883[ML16274A289], dated Sept. 29, 2016, Item 3.3.00.02a.ii.a [Index Number 764] ND 0621 [ML18131A218], dated May 9, 2018, and Item 3.3.00.03a [Index Number 777] ND 0618[ML18127A009], dated May 4, 2018. This resubmittal supersedes ND-16-1883, ND 0621 and ND-18-0618 in their entirety.

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,

'Ci/li Michael J. Yox Regulatory Affairs Director Vogtle 3&4

U.S. Nuclear Regulatory Commission ND-19-1348 Page 2 of 4

Enclosure:

Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC 3.3.00.02a.l.a [Index Number 760]

MJY/GJLVsfr

U.S. Nuclear Regulatory Commission ND-19-1348 Page 3 of 4 To:

Southern Nuclear Operating Company/ Georgia Power Company Mr. Peter P. Sena III (w/o enclosures)

Mr. D. L. McKlnney (w/o enclosures)

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

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

Mr. G. Chick Mr. M. Page Mr. P. Martino Mr. M. J. Yox Mr. A. S. Parton Ms. K. A. Roberts Mr. T. G. Petrak Mr. C. T. Defnall Mr. C. E. Morrow Mr. J. L. Hughes Mr. S. Leighty Ms. A. C. Chamberlain Mr. J. C. Haswell Document Services RTYPE: VND.LI.L06 File AR.01.02.06 cc; Nuclear Reoulatorv Commission Mr. W.Jones(w/o enclosures)

Mr. F. D. Brown Mr. C. P. Patel Mr. G. J. Khouri Ms. S. E. Temple Mr. N. D. Karlovich Mr. A. Lerch Mr. C. J. Even Mr. B. J. Kemker Ms. N. C. Coovert Mr. C. Welch Mr. J. Gaslevic Mr. V. Hall Mr. G. Armstrong Ms. T. Lamb Mr. M. Webb Mr. T. Fredette Mr. C. Weber Mr. S. Smith Mr. C. Santos Oalethorpe Power Corporation Mr. R. B. Brinkman Mr. E. Rasmussen

U.S. Nuclear Regulatory Commission ND-19-1348 Page 4 of 4 Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. S. M. Jackson Dalton Utilities Mr. T. Bundros Westinqhouse 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. Z. 8. Harper Mr. J. L. Coward Other Mr. J. E. Hosier, Bechtel Power Corporation Ms. L. Matis, TetraTech NUS, Inc.

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

Mr. 8. Roetger, Georgia Public Service Commission Ms. 8. W. Kernlzan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. 8. Blanton, Balch BIngham

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 1 of 10 Southern Nuclear Operating Company ND-19-1348 Enclosure Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC Item 3.3.00.02a.i.a [Index No. 760]

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 2 of 10 ITAAC Statement Design Commitment 2.a) The nuclear island structures, including the critical sections listed in Table 3.3-7, are seismic Category i and are designed and constructed to withstand design basis loads as specified in the Design Description, without loss of structural integrity and the safety-related functions.

3.) Walls and floors of the nuclear island structures as defined on Table 3.3-1 except for designed openings or penetrations, provide shielding during normal operations.

I nspections/Tests/Analvses I) An inspection of the nuclear island structures will be performed. Deviations from the design due to as-built conditions will be analyzed for the design basis loads, and for radiation shielding.

Acceptance Criteria i.a) A report exists which reconciles deviations during construction, including Table 3.3-1 wall and floor thicknesses, and concludes that the as-built containment internal structures, including the critical sections, conform to the approved design and will withstand the design basis loads specified in the Design Description without loss of structural integrity or the safety-related functions, and without impacting compliance with the radiation protection licensing basis.

ITAAC Completion Description Multiple ITAAC are performed to demonstrate that the nuclear island (Nl) structures, including the critical sections listed in VEGP Unit 3&4 Combined License (COL) Appendix C (Reference 1 and 2)Table 3.3-7(Attachment A), are seismic Category I and are designed and constructed to withstand design basis loads as specified in the VEGP Unit 3&4 COL Appendix C Section 3.3 Design Description, without loss of structural integrity and the safety-related functions. In addition, multiple ITAAC are performed on walls and floors of the Nl structures as defined on Table 3.3-1 (Attachment B) except for designed openings or penetrations, that provide radiation shielding during normal operations.

The subject ITAAC verifies inspection of the as-built containment internal structures, including the critical sections and Table 3.3-1 wall and floor thicknesses, and reconciles deviations during construction to the approved design such that the as-built structures will withstand design basis loads without loss of structural Integrity or the safety-related functions, and without impacting compliance with the radiation protection licensing basis.

Design bases loads are defined in VEGP Unit 3&4 COL Appendix C Section 3.3 as those loads associated with:

• Normal plant operation (including dead loads, live loads, lateral earth pressure loads, and equipment loads, including hydrodynamic loads, temperature and equipment vibration);

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 3 of 10

• External events (including rain, snow,flood, tornado, tornado generated missiles and earthquake): and

• Internal events (including flood, pipe rupture, equipment failure, and equipment failure generated missiles).

VEGP 3&4 Updated Final Safety Analysis Report(Reference 3), Section 3.7 "Seismic Design",

Section 3.8 "Design of Category I Structures", and Appendix 3H "Auxiliary and Shield Building Critical Sections" describe the analyses for the design basis loads for the Ml Structures. Section 3.8 specifies the applicable codes and standards governing the design, materials, fabrication, construction inspection and testing for the Ml structures. Section 3.8 also describes the as-built design summary reports which document that the seismic Category I structures meet the specified acceptance criteria.

Radiation zone and equipment qualification requirements are met in accordance with VEGP 3&4 UFSAR Tier 2 design criteria including UFSAR Subsections 3.11.4 "Estimated Radiation and Chemical Environment," 3D.5.1.2 "Radiation Dose," and 12.3.2.1 "Shielding, Design Objectives".

The containment internal structures, including the critical sections, listed in Attachment A, and walls and floors of the Nl structures as defined on Table 3.3-1 (Attachment B) except for designed openings or penetrations, provide radiation shielding during normal operations and are constructed as designed and specified in the VEGP Unit 3&4 COL Appendix C Section 3.3 Design Description to withstand the Design Description design basis loads without loss of structural integrity and the safety-related functions, and without impacting compliance with the radiation protection licensing basis.

The containment internal structures, including the critical sections, listed in Attachment A, and walls and floors of the Nl structures as defined on Table 3.3-1 (Attachment B), except for designed openings or penetrations, which provide radiation shielding during normal operations are inspected during construction to verify the as-built structures conform to the specified design, codes and standards. Construction identified structural deviations are documented, evaluated, and reconciled by engineering to confirm the structures' ability to withstand design basis loads without impacting compliance with the radiation protection licensing basis. The As-Built Summary Reports (References 4 through 9) exist and document the reconciliation of Nl structural deviations identified during construction and conclude that the as-built containment internal structures, including the critical sections and walls and floors of the Nl structures as defined on Table 3.3-1 except for designed openings or penetrations, will withstand the design basis loads specified in the Design Description without loss of structural integrity or the safety-related functions, and without impacting compliance with the radiation protection licensing basis.

Unit 3 & 4 Principle Closure Documents (References 4 through 9) are available for NRG inspection as part of the Unit 3 & 4 ITAAC 3.3.00.02a.i.a Completion Packages (References 10 and 11)

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 4 of 10 List of ITAAC Findings In accordance with plant procedures for ITAAC completion, Southern Nuclear Operating Company(SNC) performed a review of all findings pertaining to the subject ITAAC and associated corrective actions. This review found sixteen (16) NRC findings associated with this ITAAC.

1.05200025/2015002-01 (Closed - ML16032A554)

2.05200026/2015002-01 (Closed - ML16032A554)

3.05200025/2014005-01 (Closed - ML16032A554)

4.05200025/2012004-01 (Closed - ML13312A316)

5.05200025/2014004-01 (Closed - ML14311A666)

6.05200025/2014004-02(Closed - ML14311A666)

7.05200025/2016004-01 (Closed - ML18317A395)

8.05200026/2016004-01 (Closed - ML18317A395)

9.05200026/2017001-01 (Closed - ML17132A345)

10.05200025/2012004-01 (Closed - ML12319A458)

11. 99901439/2015-201-01 (Closed - ML18100A857)

12. 99901439/2015-201-02(Closed - ML18100A857)

13. 99901439/2014-201-01 (Closed - ML15175A446)

14. 99901425/2014-201-01 (Closed - ML18101A168)

15. 99901419/2012-201-03(Closed - ML18131A260)

16. 99901409/2011 -201-03(Closed - ML18186A573)

The IT/\AC completion review is documented in the ITAAC Completion Packages for ITAAC 3.3.00.02a.i.a Unit 3 and Unit 4(Reference 10 and 11) and are available for NRC review.

References(available for NRC inspection)

1. VEGP Unit 3 COL Appendix C, Amendment 167

2. VEGP Unit 4 COL Appendix C, Amendment 165

3. VEGP 3&4 UFSAR, Revision 8.1

4. As-Built Summary Report for Unit 3 Containment Structural Modules, SV3-AAA-BBB-###

5. As-Built Summary Report for Unit 4 Containment Structural Modules, SV4-A/\A-BBB-###

6. As-Built Summary Report for Unit 3 Other Containment Internal Structures, SV3-CCC-DDD-###

7. As-Built Summary Report for Unit 4 Other Containment Internal Structures, SV4-CCC-DDD-###

8. As-Built Summary Report for Unit 3 Nuclear Island Basemat, SV3-EEE-FFF-###

9. As-Built Summary Report for Unit 4 Nuclear Island Basemat, SV4-EEE-FFF-###

10. 3.3.00.02a.i.a-U3-CP-Rev0, IT/\AC Completion Package

11. 3.3.00.02a.i.a-U4-CP-Rev0, IT/VAC Completion Package

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

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 5 of 10 Attachment A: Excerpt of COL Appendix C Table 3.3-7 Table 3.3-7 Nuclear Island Critical Structural Sections Containment Internal Structures South west wall of the refueling cavity South wall of the west steam generator compartment North east wall of the in-containment refueling water storage tank In-containment refueling water storage tank steel wall Column supporting the operating floor

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 6 of 9 Attachment B: Excerpt of COL Appendix C Table 3.3-1 Table 3.3-1 Definition of Waii Thicknesses for Nuclear island Buildings,Turbine Building, and Annex Buiiding^^)

Applicable Radiation Shielding Floor Elevation or Concrete Wall Wail or Section Description Column Lines^^) Elevation Range^'^^' (Yes/No)

Containment Building Internal Structure^^^^

Shield Wall between Reactor E-W wall parallel with From 71'-6" to 83'-0" 3'-0"d°) Yes Vessel Cavity and RCDT Room column line 7(Inside face is 3'-0" north of column line 7. Width of wall section with stated thickness is defined by inside wall of reactor vessel cavity.)

West Reactor Vessel Cavity Wall N-S wall parallel with From 83'-0" to 98'-0" 7'-6"(io) Yes column line N (Width of wall section with stated thickness is defined by inside wall of reactor vessel cavity.)

North Reactor Vessel Cavity E-W wall parallel with From 83'-0" to 98'-0" 9'-0"(io) Yes Wall column line 7(Width of wall section with stated thickness is defined by inside wall of reactor vessel cavity.)

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 7 of 9 Attachment B: Excerpt of COL Appendix C Table 3.3-1 Table 3.3-1 Definition of Waii Thicknesses for Nuclear Island Buildings, Turbine Building, and Annex Building^^^

Applicable Radiation Shielding Floor Elevation or Concrete Wall Wall or Section Description Column Lines^^) Elevation Range^)<> (Yes/No)

Containment Building Internal Structure^^^

East Reactor Vessel Cavity Wall N-S wall parallel with From 83'-0" to 98'-0" Yes column line N (Width of wall section with stated thickness is defined by inside wall of reactor vessel cavity.)

West Refueling Cavity Wall N-S wall parallel with From 98'-0"to135'-3" 4'-0" Yes column line N North Refueling Cavity Wall E-W wall parallel with From 98'-0" to 135'-3" 4'-0" Yes column line 7 East Refueling Cavity Wall N-S wall parallel with From 98"-0"to135'-3" 4'-0" Yes column line N South Refueling Cavity Wall E-W wall parallel with From 98'-0"to135'-3" 4'-0" Yes column line 7 South wall of west steam Not Applicable From103'-0" to153'-0" 2'-6" Yes generator compartment West wall of west steam N-S wall parallel with From 103'-0" to 153'-0" 2'-6" Yes generator compartment column line N North wall of west steam Not Applicable From 103'-0" to 153'-0" 2'-6" Yes generator compartment South wall of pressurizer Not Applicable From103'-0"to153'-6" 2'-6" Yes compartment West wall of pressurizer N-S wall parallel with From 107'-2"to 160'-0" 2'-6" Yes compartment column line N

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 8 of 9 Attachment B: Excerpt of COL Appendix C Table 3.3-1 Table 3.3-1 Definition of Wall Thicknesses for Nuclear Island Buildings, Turbine Building, and Annex Building^^'

Applicable Radiation Shielding Floor Elevation or Concrete Wall Wall or Section Description Column Llnes^^) Elevation Range<^><> (Yes/No)

Containment Building Internal Structure^^^^

North wall of pressurizer E-W wall parallel with From 107'-2" to 160'-0" 2'-6" Yes compartment column line 7 East wall of pressurizer N-S wall parallel with From118'-6" to160'-0" 2'-6" Yes compartment column line N North-east wall of In-contalnment Parallel to column line N From 103'-0"to 135'-3" 2'-6" No refueling water storage tank West wall of In-contalnment Not Applicable From 103'-0" to 135'-3" 5/8" steel plate with No refueling water storage tank stiffeners South wall of east steam Not Applicable From 87'-6"to153'-0" 2'-6" Yes generator compartment East wall of east steam N-S wall parallel with From 94'-0" to 153'-0" 2'-6" Yes generator compartment column line N North wall of east steam Not Applicable From 87'-6"to153'-0"wlth 2'-6" Yes generator compartment a 158'-0" portion

U.S. Nuclear Regulatory Commission ND-19-1348 Enclosure Page 9 of 9 Attachment B: Excerpt of COL Appendix C Table 3.3-1

1. The column lines and floor elevations are identified and included on Figures 3.3-1 through 3.3-13.

2. These wall (and floor) thicknesses have a construction tolerance of +/- 1 inch, except as noted and for exterior walls below grade where the tolerance is +12 inches, -1 inch. These tolerances are not applicable to the nuclear island basemat.

3. For walls that are part of structural modules, the concrete thickness also includes the steel face plates. Where faceplates with a nominal thickness of 0.5 inches are used in the construction of the wall modules, the wall thicknesses in this column apply. Where faceplates thicker than the nominal 0.5 inches are used in the construction of the structural wall modules, the wall thicknesses in the area of the thicker faceplates are greater than indicated in this column by the amount of faceplate thickness increase over the nominal 0.5 inches. Overlay plates are not considered part of the faceplates, and thus are not considered in the wall thicknesses identified in this column.

4. For floors with steel surface plates, the concrete thickness also includes the plate thickness.

5. Where a wall (or a floor) has openings, the concrete thickness does not apply at the opening.

6. N/A to ITAAC 3.3.00.02a.i.a.

7. The Wall or Section Description, Column Line information, and Floor Elevation or Elevation Ranges are provided as reference points to define the general location. The concrete thickness of an item intersecting other walls, roofs or floors at a designated location (e.g., column line) is not intended to be measured to the stated column line, but only to the point where the intersection occurs.

8. Where applicable, the upper wall portions extend to their associated roofs, which may vary in elevation, e.g., sloped roofs.

9. From one wall/floor section to another, the concrete thickness transitions from one thickness to another, consistent with the configurations in Figures 3.3-1 through 3.3-14.

10. N/A to ITAAC 3.3.00.02a.i.a.

11. N/A to \TAAC 3.3.00.02a.i.a.

12. N/A to ITAAC 3.3.00.02a.i.a.

13. N/A to ITAAC 3.3.00.02a.i.a.

14. N/A to IT/\AC 3.3.00.02a.i.a.

15. Reconciliation of construction deviations in the nuclear island structures from the thickness and tolerances specified in this table is included in the reconciliation reports, and demonstrate that the as-built structures will withstand design basis loads without loss of structural integrity or safety functions and without impacting compliance with the radiation protection licensing basis, such as GDC 19, established radiological zoning and equipment qualification in accordance with ITAAC 3.3.00.02a.i.a, 3.3.00.02a.i.b, 3.3.00.02a.i.c, or 3.3.00.02a.i.d.

16. N/A to ITAAC 3.3.00.02a.i.a.

17. N/A to ITAAC 3.3.00.02a.i.a.

18. Nonconformances from the thicknesses and tolerances specified in Table 3.3-1 (i.e. out of tolerance conditions) are addressed under the 10 CFR Part 50, Appendix B process and subsequently are screened in accordance with the 10 CFR Part 52, Appendix D, Section VIII process, to ensure that the licensing basis is adequately maintained. Construction deviations will continue to be assessed against the licensing basis requirements and will be addressed in accordance with licensee procedures and regulatory requirements and, if applicable, a license amendment will be obtained prior to implementation of the change.