ND-23-1001, ITAAC Closure Notification on Completion of ITAAC 2.7.01.05.i (Index Number 684)

ITAAC Closure Notification on Completion of ITAAC 2.7.01.05.i (Index Number 684)

ML23005A152
Person / Time
Site:	Vogtle
Issue date:	01/05/2023
From:	Coleman J Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
ND-23-1001
Download: ML23005A152 (1)

Text

Jamie M, Coleman 7825 River Road A Southern Nuclear Regulatory Affairs Director Waynesboro, GA 30830 Vogtle 3 & 4 706-848-6926 tel

January 5, 2023

Docket No.: 52-026

ND-23-1001 10 CFR 52.99(c)(1)

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

Southern Nuclear Operating Company Vogtle Electric Generating Plant Unit 4 ITMC Closure Notification on Completion of ITMC 2.7.01.05.i [Index Number 684]

Ladies and Gentlemen:

In accordance with 10 CFR 52.99(c)(1 ), the purpose of this letter is to notify the Nuclear Regulatory Commission (NRG) of the completion of Vogtle Electric Generating Plant (VEGP) Unit 4 Inspections, Tests, Analyses, and Acceptance Criteria ITMC Item 2.7.01.05.i [Index Number 684]. This ITMC confirms that the seismic Category I equipment identified in VEGP Combined License (COL)

Appendix C Table 2.7.1-1 for the Nuclear Island Nonradioactive Ventilation System (VBS) can withstand seismic design basis loads without loss of safety function. The closure process for this IT MC is based on the guidance described in NEI 08-01, "Industry Guideline for the IT MC Closure Process under 10 CFR Part 52," which was endorsed by the NRG in Regulatory Guide 1.215.

This letter contains no new NRG regulatory commitments. Southern Nuclear Operating Company requests NRG staff confirmation of this determination and publication of the required notice in the Federal Register per 10 CFR 52.99.

If there are any questions, please contact Kelli Roberts at 706-848-6991.

Respectfully submitted,

'ij~ ~

Jamie M. Coleman Regulatory Affairs Director Vogtle 3 & 4

Enclosure : Vogtle Electric Generating Plant (VEGP) Unit 4 Completion of ITMC 2.7.01.05.i [Index Number 684]

JMC/AD/sfr U.S. Nuclear Regulatory Commission ND-23-1001 Page 2 of 2

cc: Regional Administrator, Region II Director, Office of Nuclear Reactor Regulation (NRR)

Director, Vogtle Project Office NRR Senior Resident Inspector-Vogtle 3 & 4 U.S. Nuclear Regulatory Commission ND-23-1001 Enclosure Page 1 of 6

Southern Nuclear Operating Company ND-23-1001 Enclosure

Vogtle Electric Generating Plant (VEGP) Unit 4 Completion of ITAAC 2.7.01.05.i [Index Number 684]

U.S. Nuclear Regulatory Commission ND-23-1001 Enclosure Page 2 of 6

IT AAC Statement

Design Commitment

5. The seismic Category I equipment identified in Table 2.7.1-1 can withstand seismic design basis loads without loss of safety function.

Inspections, Tests, Analyses

i) Inspection will be performed to verify that the seismic Category I equipment identified in Table 2.7.1-1 is located on the Nuclear Island.

ii) Type tests, analyses, or a combination of type tests and analyses of seismic Category 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 by the tested or analyzed conditions.

Acceptance Criteria

i) The seismic Category I equipment identified in Table 2.7. 1-1 is located on the Nuclear Island.

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

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

ITAAC Determination Basis

This ITMC requires that inspections, tests, and analyses be performed and documented to ensure the Nuclear Island Nonradioactive Ventilation System (VBS) equipment identified as seismic Category I in the Combined License (COL) Appendix C, Table 2.7.1-1 was designed and constructed in accordance with applicable requirements.

i) The seismic Category I equipment identified in Table 2.7.1-1 is located on the Nuclear Island.

To assure that seismic Category I equipment can withstand seismic design basis loads without loss of safety function, all the equipment in Table 2.7.1-1 was designed to be located on the seismic Category I Nuclear Island. In accordance with Equipment Qualification (EQ) ITAAC As Built Walkdown Guideline (Reference 1) and EQ IT MC As-built Installation Documentation Guideline (Reference 2), inspections were conducted of the VBS to confirm the satisfactory installation of the seismically qualified equipment. The inspections included verification of equipment make/model/serial number and verification of equipment location (Building, Elevation, Room). The EQ As-Built Reconciliation Report (EQRR) (Reference 3) identified in Attachment A documents the results of the inspections and concludes that the seismic Category I equipment is located on the Nuclear Island.

U.S. Nuclear Regulatory Commission ND-23-1001 Enclosure Page 3 of 6

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

Seismic Category I equipment in Table 2.7.1-1 required type tests and/or analyses to demonstrate structural integrity and operability. Structural integrity of the seismic Category I valves was demonstrated by analysis in accordance with American Society of Mechanical Engineers (ASME) Code Section Ill (Reference 4 ). Functionality of the subset of active safety related valves under seismic loads was determined using the guidance of ASME QME-1-2007 (Reference 5).

Safety-related (Class 1 E) electrical equipment in the Table is seismically qualified by type testing combined with analysis in accordance with Institute of Electrical and Electronics Engineers (IEEE) Standard 344-1987 (Reference 6). This equipment includes safety-related active valve accessories such as electric actuators, position switches, pilot solenoid valves and electrical connector assemblies.

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 about the methods used to qualify AP1000 safety-related equipment is provided in the Updated Final Safety Analysis Report (UFSAR) Appendix 3D (Reference 7). The EQ Reports (Reference 8) identified in Attachment A contain applicable test reports and associated documentation and conclude that the seismic Category I equipment can withstand seismic design basis loads without loss of safety function.

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

Inspections (Reference 1 and Reference 2) were conducted to confirm the satisfactory installation of the seismically qualified equipment in Table 2.7.1-1. The inspections verified 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 was given to the definition of the clearances needed around the equipment mounted in the plant to permit the equipment to move during a postulated seismic event without causing impact between adjacent pieces of safety related equipment. When required, seismic testing measuring the maximum dynamic relative displacement of the top and bottom of the equipment was performed. EQ Reports (Reference 8) 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 were defined based on the test configuration and/or other design requirements.

Attachment A identifies the EQRR (Reference 3) completed to verify that the as-built seismic Category I equipment listed in Table 2.7.1-1, including anchorage, is seismically bounded by the tested or analyzed conditions, IEEE Standard 344-1987 (Reference 6), and NRC Regulatory Guide 1.100 (Reference 9).

U.S. Nuclear Regulatory Commiss ion ND-23-1001 Enclosure Page 4 of 6

Together, these reports (References 3 and 8) provide evidence that the ITAAC Acceptance Criteria requirements are met:

• The seismic Category I equipment identified in Table 2.7.1-1 is 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 3 and 8 are available for NRC inspection as part of the Unit 4 ITAAC 2.7.01.05.i Completion Package (Reference 10).

ITAAC Finding Review

In accordance with plant procedures for ITAAC completion, Southern Nuclear Operating Company (SNC) performed a review of all IT AAC findings pertaining to the subject IT AAC and associated corrective actions. This finding review, which included now-consolidated IT AAC Index 685 and 686, found the following relevant IT AAC finding associated with this IT AAC.

1. NON 99901412/2012-201-02 (Closed)

ITAAC Completion Statement

Based on the above information, SNC hereby notifies the NRC that ITAAC 2.7.01.05.i was performed for VEGP Unit 4 and that the prescribed acceptance criteria are met.

Systems, structures, and components verified as part of this ITAAC are being maintained in their as-designed, ITAAC compliant condition in accordance with approved plant programs and procedures.

References (available for NRC inspection)

1. ND-RA-001-014, "EQ ITAAC As-Built Walkdown Guideline", Version 3.1 2. ND-RA-001-016, "EQ ITAAC As-built Installation Documentation Guideline" Version 1.0

3. EQ As-Built Reconciliation Report (EQRR) as identified in Attachment A

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

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

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

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

7. Vogtle 3&4 Updated Final Safety Analysis Report Appendix 3D, "Methodology for Qualifying AP1000 Safety-Related Electrical and Mechanical Equipment", Revision 11.1

8. Equipment Qualification (EQ) Reports as identified in Attachment A U.S. Nuclear Regulatory Commission ND-23-1001 Enclosure Page 5 of 6

9. Regulatory Guide 1. 100, Rev. 2, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants"

10. 2.7.01.05.i-U4-CP-Rev0, ITMC Completion Package U.S. Nuclear Regulatory Commission ND-23-1001 Enclosure Page 6 of 6

Attachment A

System: Nuclear Island Nonradioactive Ventilation System

Equipment Tag No.+ Seismic Type of EQ Reports As-Built EQRR Name+ Cat. I+ Qual. (Reference 8) (Reference 3)

MCR Supply VBS-PL-V186* Yes Type Test SV4-PV11-VBR-006 / 2.7.01.05.i-U4-EQRR-Air Isolation &Analysis SV4-PV11-VBR-005 PCD001-Rev 0 Valve MCR Supply Type Test SV4-PV11-VBR-006 / 2.7.01.05.i-U4-EQRR-Air Isolation VBS-PL-V187* Yes SV4-PV11-VBR-005 PCD001-Rev 0 Valve &Analysis

MCR Return Type Test SV4-PV11-VBR-006 / 2.7.01.05.i-U4-EQRR-Air Isolation VBS-PL-V188* Yes &Analysis SV4-PV11-VBR-005 PCD001-Rev 0 Valve

MCR Return Type Test SV4-PV11-VBR-006 / 2. 7.01.05.i-U4-EQRR-Air Isolation VBS-PL-V189* Yes &Analysis SV4-PV11-VBR-005 PCD001-Rev 0 Valve

MCR Exhaust Type Test SV4-PV11-VBR-006 / 2.7.01.05.i-U4-EQRR-Air Isolation VBS-PL-V 190* Yes &Analysis SV4-PV11-VBR-005 PCD001 - Rev 0 Valve

MCR Exhaust Type Test SV4-PV11 -VBR-006 / 2. 7.01.05.i-U4-EQRR-Air Isolation VBS-PL-V191

• Yes &Analysis SV4-PV11-VBR-005 PCD001-Rev 0 Valve PWS MGR Type Test SV4-PV02-VBR-010 / 2. 7.01.05.i-U4-EQRR-Isolation Valve PWS-PL-V418 Yes &Analysis SV4-PV02-VBR-009 PCD001-Rev 0 PWS MGR PWS-PL-V420 Yes Type Test SV4-PV02-VBR-010 / 2. 7.01.05.i-U4-EQRR-Isolation Valve &Analysis SV4-PV02-VBR-009 PCD001-Rev 0 PWS MGR PWS-PL-V498 Yes Type Test SV4-PV18-VBR-002 2. 7.01.05.i-U4-EQRR-Vacuum Relief &Analysis /SV4-PV18-VBR 001 PCD001-Rev 0 MCRSDS Type Test SV4-PV11-VBR-006 2.7.01.05.i-U4-EQRR-(Vent) SDS-PL-V001* Yes &Analysis / SV4-PV11-VBR-005 PCD001-Rev 0 Isolation Valve MGRSDS Type Test SV4-PV11-VBR- 006 / 2. 7.01.05.i-U4-EQRR-(Vent) SDS-PL-V002* Yes &Analysis SV4-PV11-VBR-005 PCD001-Rev 0 Isolation Valve MGRWWS WWS-PL-V506 SV4-PV03-VBR- 006 / 2.7.01.05.i-U4-EQRR-Isolation Valve Yes Analysis SV4-PV03-VBR-005 PCD001-Rev 0

+ Excerpt from COL Appendix C, Table 2.7.1-1

• Indicates 1 E component, per Table 2.7.1-1

Attachment A Page 1 of 1