ND-18-0257, Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.3.07.05.i (Index Number 396)

Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.3.07.05.i (Index Number 396)

ML18054A369
Person / Time
Site:	Vogtle
Issue date:	02/22/2018
From:	Yox M Southern Nuclear Operating Co
To:	Document Control Desk, Office of New Reactors
References
ITAAC 2.3.07.05.i, ND-18-0257
Download: ML18054A369 (10)

Text

Michael J. Yox 7825 River Road Regulatory Affairs Director Waynesboro, GA 30830

^ Southern Nuclear Vogtle 3 & 4 706-848-6459 tei 410-474-8587 ceii F£B 2 2 2018 myox @southernco.com Docket Nos.: 52-025 52-026 ND-18-0257 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 2.3.07.05.1 [Index Number 3961 Ladies and Gentlemen:

Pursuant to 10 CFR 52.99(c)(3), Southern Nuclear Operating Company hereby notifies the NRC that as of February 16, 2018, Vogtle Electric Generating Plant (VEGP) Unit 3 and Unit 4 Uncompleted Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Item 2.3.07.05.1

[Index Number396] has not been completed greater than 225-days prior to 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 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.

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 2.3.07.05.1 [Index Number 396]

MJY/KJD/amw

U.S. Nuclear Regulatory Commission ND-18-0257 Page 2 of 3 To:

Southern Nuclear Operating Company/ 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. McKlnney Mr. M. J. Yox Mr. D. L. Fulton Mr. J. D. Williams 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. C. Chamberlain Mr. D. R. Culver Mr. T. G. Petrak Document Services RTYPE: VND.LI.L06 File AR.01.02.06 cc:

Nuclear Reaulatorv 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. P. B. Donnelly 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 Oalethorpe Power Corporation Mr. R. B. Brinkman Municipal Electric Authorltv of Georgia Mr. J. E. Fuller Mr. S. M. Jackson

U.S. Nuclear Regulatory Commission ND-18-0257 Page 3 of 3 Dalton Utilities Mr. T. Bundros Westinqhouse Electric Company. LLC Dr. L. Oriani (w/o enclosures)

Mr. D. C. 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., CDS 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-18-0257 Enclosure Page 1 of 7 Southern Nuclear Operating Company ND-18-0257 Enclosure Vogtle Electric Generating Plant (VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC 2.3.07.05.1 [Index Number 396]

U.S. Nuclear Regulatory Commission ND-18-0257 Enclosure Page 2 of 7 ITAAC Statement Desion Commitment:

5. The seismic Category I components identified in Table 2.3.7-1 can withstand seismic design basis loads without loss of safety functions.

Insoections. Tests. Analyses:

i) Inspection will be performed to verify that the seismic Category I components identified in Table 2.3.7-1 are located on the Nuclear Island.

ii) Type tests, analyses, or a combination of type tests and analyses of seismicCategory I equipment will be performed.

iii) Inspection will be performed for the existence of a report verifying that the as-built equipment including anchorage is seismically bounded bythe tested or analyzed conditions.

Acceotance Criteria:

i) Theseismic Category Icomponents identified in Table 2.3.7-1 are located on the Nuclear Island.

ii) Areport exists and concludes that the seismic Category Iequipment can withstand seismic design basis loads without loss of safety function.

iii) Areport exists and concludes that the as-built equipment including anchorage isseismically bounded by the tested or analyzed conditions.

ITAAC Completion Description This ITAAC requires that inspections, tests, and analyses be performed and documented to ensurethe Spent Fuel Pool Cooling System (SFS) components (equipment) identified as seismic Category Iin the Combined License (COL) Appendix C, Table 2.3.7-1 (the Table) are designed and constructed in accordance with applicable requirements.

ii The seismic Cateoorv I components identified in Table 2.3.7-1 are located on the Nuclear Island.

To assure that seismic Category Icomponents can withstand seismic design basis loads without lossof safety function, all the components in the Table are designed to be located on the seismic Category I Nuclear Island. In accordance with Equipment Qualification (EG)

Walkdown ITAAC Guideline (Reference 1), an inspection is conducted of the SFS to confirm the satisfactory installation of the seismically qualified components. The inspection includes verification of component make/model/serial number and verification of component location (Building, Elevation, Room). The EG As-Built Reconciliation Reports (EQRR) (Reference 2) identified in Attachment A document the results of the inspection and conclude that the seismic Category Icomponents are located on the Nuclear Island.

U.S. Nuclear Regulatory Commission ND-18-0257 Enclosure Page 3 of 7 ill A report exists and concludes that tfie seismic Category I eouioment can withstand seismic desion basis loads without loss of safety function.

Seismic Category I equipment in the Table requires type tests and/or analyses to demonstrate structural integrity and operability. Structural integrity of the seismic Category I valves is demonstrated by analysis in accordance with American Society of Mechanical Engineers (ASME) Code Section III (Reference 3). Functionality of the subset of active safety-related valves under seismic loads is determined using the guidance of ASME QME-1-2007 (Reference 4).

Safety-related (Class 1E) electrical equipment in the Table is seismically qualified bytype testing combinedwith analysis in accordance with Institute of Electrical and Electronics Engineers (IEEE) Standard 344-1987 (Reference 5). This equipment includes safety-related (Class 1E) field sensors and the safety-related active valve accessories such as electric actuators, position switches, pilot solenoid valves and electrical connectorassemblies. The specific qualification method (i.e., type testing, analysis, or combination) used for each piece of equipment in the Table is identified in Attachment A. Additional information aboutthe methods used to qualify API 000 safety-related equipment is provided in the Updated Final Safety Analysis Report (UFSAR) Appendix 3D (Reference 6). The EQ Reports (Reference 7) identified in Attachment Acontain applicable test reports and associated documentation and conclude that the seismic Category Iequipment can withstand seismic design basis loadswithout loss of safety function.

iiil A reoort exists and concludes that the as-built eouioment includino anchoraoe is seismicallv bounded bv the tested or analvzed conditions.

An inspection (Reference 1) is conducted to confirm the satisfactory installation ofthe seismically qualified equipment in the Table. The inspection verifies the equipment make/model/serial number, as-designed equipment mounting orientation, anchorage and clearances, and electrical and other interfaces. The documentation of installed configuration of seismically qualified equipment includes photographs and/or sketches/drawings of equipment/mounting/interfaces.

As part of the seismic qualification program, consideration is given to the definition of the clearances needed around the equipment mounted in the plantto permit the equipmentto move during a postulated seismic event without causing impact between adjacent pieces of safety-related equipment. This isdone as part of seismic testing by measuring the maximum dynamic relative displacement of thetop and bottom of the equipment. EQ Reports (Reference 7) identify the equipment mounting employed for qualification and establish interface requirements for assuring that subsequent in-plant installation does not degrade the established qualification.

Interface requirements are defined based on the test configuration and otherdesign requirements.

Attachment A identifies the EQRR (Reference 2) completed to verify that the as-built seismic Category Iequipment listed in the Table, including anchorage, are seismically bounded by the tested or analyzed conditions, IEEE Standard 344-1987 (Reference 5), and NRG Regulatory Guide 1.100 (Reference 8).

U.S. Nuclear Regulatory Commission ND-18-0257 Enclosure Page 4 of 7 Together, these reports (References 2 and 7) provide evidence that the ITAAC Acceptance Criteria requirements are met:

• The seismic Category I components identified in Table 2.3.7-1 are located on the Nuclear Island;

• A report exists and concludes that the seismic Category I equipment can withstand seismic design basis loads without loss of safety function; and

• A report exists and concludes that the as-built equipment including anchorage is seismically bounded by the tested or analyzed conditions.

References 2 and 7 are available for NRC inspection as part of the Unit 3 and Unit 4 ITAAC 2.3.07.05.1 Completion Packages (References 9 and 10, respectively).

List of ITAAC Findings In accordance with plant procedures for ITAAC completion. Southern NuclearOperating Company (SNC) performed a review of all ITAAC findings pertaining to the subject ITAAC and associated corrective actions. This finding review, which included now-consolidated ITAAC Indexes 397 and 398, found one relevant ITAAC finding associated with this ITAAC.

1) Notice of Nonconformance (NGN) 99901412/2012-201 -02 (Closed)

References (available for NRC mspectlon)

1. ND-xx-xx-001, "EG Walkdown IT/\AC Guideline"

2. EG As-Built Reconciliation Reports (EQRR) as identified in Attachment Afor Units 3 and 4

3. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,Section III, "Rules for Construction of Nuclear Power Plant Components," 1998 Edition with 2000 Addenda

4. ASME QME-1 -2007, "Qualification of Active Mechanical Equipment Used in Nuclear Power Plants," The American Society of Mechanical Engineers, June 2007

5. IEEE Standard 344-1987, "IEEE Recommended Practices for Seismic Qualification of Class 1E Equipmentfor Nuclear Power Generating Stations"

6. Vogtle 3&4 Updated Final Safety Analysis Report Appendix 3D, "Methodology for Qualifying API 000 Safety-Related Electrical and Mechanical Equipmenf

7. Equipment Qualification (EQ) Reports as identified inAttachment A

8. Regulatory Guide 1.100, Rev. 2, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants"

U.S. Nuclear Regulatory Commission ND-18-0257 Enclosure Page 5 of 7

9. 2.3.07.05.i-U3-CP-Rev X, "Completion Package for Unit 3 ITAAC 2.3.07.05.1 [Index Number 396]"

10. 2.3.07.05.i-U4-CP-Rev X, "Completion Package for Unit 4 ITAAC 2.3.07.05.1 [Index Number 396]"

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

U.S. Nuclear Regulatory Commission ND-18-0257 Enclosure Page 6 of 7 Attachment A System: Spent Fuel Pool Cooling System (SFS)

Seismic As-Buiit EQ EQRR Equipment Name

• Tag No.

• Type of Qual. EQ Reports (Reference 7)

Cat. 1 (Reference 2)

• APP-JE52-VBR-002 / 2.3.07.05.i-U3-EQRR-Spent Fuel Pool Level Sensor SFS-019A Yes Type Test APP-JE52-VBR-001 PCDXXX APP-JE52-VBR-002 / 2.3.07.05.i-U3-EQRR-Spent Fuel Pool Level Sensor SFS-019B Yes Type Test APP-JE52-VBR-001 PCDXXX APP-JE52-VBR-002 / 2.3.07.05.i-U3-EQRR-Spent Fuel Pool Level Sensor SFS-019C Yes Type Test APP-JE52-VBR-001 PCDXXX Refueling Cavity Drain to SGS Type Test & APP-PV11-VBR-002 / 2.3.07.05.i-U3-EQRR-SFS-PL-V031 Yes Compartment Isolation Vaive Analysis APP-PV11-VBR-001 PCDXXX Refueling Cavity to SFS Pump Type Test & APP-PV11-VBR-002 / 2.3.07.05.i-U3-EQRR-SFS-PL-V032 Yes Suction Isolation Valve Analysis APP-PV11-VBR-001 PCDXXX Refueling Cavity Drain to Type Test & APP-PV10-VBR-002/ 2.3.07.05.i-U3-EQRR-SFS-PL-V033 Yes Containment Sump Isolation Valve Analysis APP-PV10-VBR-001 PCDXXX IRWST to SFS Pump Suction Line Type Test & APP-PV11-VBR-002/ 2.3.07.05.i-U3-EQRR-SFS-PL-V039 Yes Isolation Valve Analysis APP-PV11-VBR-001 PCDXXX Fuel Transfer Canal to SFS Pump Type Test & APP-PV11-VBR-002/ 2.3.07.05.i-U3-EQRR-SFS-PL-V040 Yes Suction Iso. Valve Analysis APP-PV11-VBR-001 PCDXXX Cask Loading Pit to SFS Pump Type Test & APP-PV11-VBR-002 / 2.3.07.05.i-U3-EQRR-SFS-PL-V041 Yes Suction Isolation Valve Analysis APP-PV11-VBR-001 PCDXXX Cask Loading Pit to SFS Pump Type Test & APP-PV11-VBR-002 / 2.3.07.05.i-U3-EQRR-SFS-PL-V042 Yes Suction Isolation Valve Analysis APP-PV11-VBR-001 PCDXXX SFS Pump Discharge Line to Cask Type Test & APP-PV11-VBR-002/ 2.3.07.05.i-U3-EQRR-SFS-PL-V045 Yes Loading Pit Isolation Valve Analysis APP-PV11-VBR-001 PCDXXX Cask Loading Pit to WLS Isolation Type Test & APP-PV10-VBR-002/ 2.3.07.05. i-U3-EQRR-SFS-PL-V049 Yes Valve Analysis APP-PV10-VBR-001 PCDXXX Spent Fuel Pool to Cask Type Test & APP-PV10-VBR-008 / 2.3.07.05.i-U3-EQRR-SFS-PL-V066 Yes Washdown Pit Isolation Valve Analysis APP-PV10-VBR-007 PCDXXX CaskWashdown Pit Drain Isolation Type Test & APP-PV11-VBR-002 / 2.3.07.05.i-U3-EQRR-SFS-PL-V068 Yes Valve Analysis APP-PV11-VBR-001 PCDXXX

U.S. Nuclear Regulatory Commission ND-18-0257 Enclosure Page 7 of 7 Attachment A System: Spent Fuel Pool Cooling System (SFS)

Seismic As-Buiit EQ EQRR Equipment Name

• Tag No. ^ Type of Qual. EQ Reports (Reference 7)

Cat. i ^ (Reference 2)

• Refueling Cavity Drain Line Check APP-PV03-VBR-014 / 2.3.07.05.i-U3-EQRR-SFS-PL-V071 Yes Analysis Valve APP-PV03-VBR-013 PCDXXX Refueling Cavity Drain Line Check APP-PV03-VBR-014/ 2.3.07.05.i-U3-EQRR-SFS-PL-V072 Yes Analysis Valve APP-PV03-VBR-013 PCDXXX SFS Containment Floodup Isolation Type Test & APP-PV11-VBR-002/ 2.3.07.05.i-U3-EQRR-SFS-PL-V075 Yes Valve Analysis APP-PV11-VBR-001 PCDXXX

+ Excerpt from COL Appendix C Table 2.3.7-1

• The Unit 4 As-Built EQRR are numbered "2.3.07.05.i-U4-EQRR-PCDXXX"