ND-22-0981, ITAAC Closure Notification on Completion of ITAAC 2.2.03.08c.vii (Index Number 192)

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ITAAC Closure Notification on Completion of ITAAC 2.2.03.08c.vii (Index Number 192)
ML23146A153
Person / Time
Site: Vogtle Southern Nuclear icon.png
Issue date: 05/26/2023
From: Coleman J
Southern Nuclear Operating Co
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
ND-22-0981, ITAAC 2.2.03.08c.vii
Download: ML23146A153 (1)


Text

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

May 26, 2023

Docket No.: 52-026

ND-22-0981 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 ITAAC Closure Notification on Completion of ITAAC 2.2.03.08c.vii [Index Number 192]

Ladies and Gentlemen:

In accordance with 10 CFR 52.99(c)(1), the purpose of this letter is to notify the Nuclear Regulatory Commission (NRC) of the completion of Vogtle Electric Generating Plant (VEGP) Unit 4 Inspections, Tests, Analyses, and Acceptance Criteria ITAAC item 2.2.03.08c.vii [Index Number 192]. This ITAAC performed an inspection of the as-built plate located above the containment recirculation screens and verified that the plate size and location satisfy the acceptance criteria. The closure process for this ITAAC is based on 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.

This letter contains no new NRC regulatory commitments. Southern Nuclear Operating Company (SNC) requests NRC 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,

Jamie M. Coleman Regulatory Affairs Director Vogtle 3 & 4

Enclosure:

Vogtle Electric Generating Plant (VEGP) Unit 4 Completion of ITAAC 2.2.03.08c.vii [Index Number 192]

JMC/PAW/sfr

U.S. Nuclear Regulatory Commission ND-22-0981 Page 2 of 2

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

Director, Vogtle Project Office NRR Senior Resident Inspector - Vogtle 3 & 4

U.S. Nuclear Regulatory Commission ND-22-0981 Enclosure Page 1 of 3

Southern Nuclear Operating Company ND-22-0981 Enclosure

Vogtle Electric Generating Plant (VEGP) Unit 4 Completion of ITAAC 2.2.03.08c.vii [Index Number 192]

U.S. Nuclear Regulatory Commission ND-22-0981 Enclosure Page 2 of 3

ITAAC Statement

Design Commitment

8.c) The PXS provides RCS makeup, boration, and safety injection during design basis events.

Inspections, Tests, Analyses

vii) Inspection of the as-built components will be conducted for the plate located above the containment recirculation screens.

Acceptance Criteria

vii) The plate located above the containment recirculation screens is no more than 1 ft, 3 in above the top of the face of the screens and extends at least 8 ft, 3 in perpendicular to the front and at least 7 ft to the side of the face of the screens.

ITAAC Determination Basis

Multiple ITAAC are performed to demonstrate the Passive Core Cooling System (PXS) provides Reactor Coolant System (RCS) makeup, boration, and safety injection during design basis events. For this ITAAC, an inspection of the as-built components was conducted for the plate located above the containment recirculation screens (PXS-MY-Y02A and PXS -MY-Y02B) to confirm the plate is located no more than 1 ft, 3 in above the top of the face of the screens and extends at least 8 ft, 3 in perpendicular to the front and least 7 ft to the side of the face of the screens.

Inspections were performed of the as -built components by taking as -built measurements using survey equipment in accordance with Nuclear Construction and Startup Procedure (NCSP) 3 -24, Field Surveying (Reference 1). NCSP 3 -24 was used to establish layout and control points for determining the distance between the plate surface to the top and side face of the containment recirculation screens. The protective plate maximum height dimension is the distance between the top of the screens and the underside of the protective plate module top plate at the exposed edges of the protective plate, which extends into the containment recirculation water flow (east toward steam generator 2, and north toward the corridor), as described in the Updated Final Safety Analysis Report (UFSAR) Subsection 6.3.2.2.7 (Reference 2 ). Measurements were taken using survey equipment in accordance with site survey and measurement procedures to derive the distance between the as built components.

The results of the inspection are documented in the Unit 4 inspection report (Reference 3) confirming the plate is no more than 1 ft, 3 in above the top of the face of the screens (maximum as -built distance is 1 ft, 1 in), extends at least 8 ft, 3 in perpendicular to the front of the screens (minimum as-built distance 8 ft, 7.6 in), extends at least 7 ft to the side of the face of the screens (minimum as-built distance is 7 ft, 0.3 in), which meets the ITAAC Acceptance Criteria.

Reference 3 is available for NRC inspection as part of the Unit 4 ITAAC 2.2.03.08c.vii Completion Package (Reference 4).

U.S. Nuclear Regulatory Commission ND-22-0981 Enclosure Page 3 of 3

ITAAC Finding Review

In accordance with plant procedures for ITAAC completion, Southern Nuclear Operating Company (SNC) performed a review of all ITAAC findings pertaining to the subject ITAAC and associated corrective actions. This review found that there are no relevant ITAAC findings associated with this ITAAC. The ITAAC completion review is documented in the ITAAC Completion Package for ITAAC 2.2.03.08c.vii (Reference 4) and is available for NRC review.

ITAAC Completion Statement

Based on the above information, SNC hereby notifies the NRC that ITAAC 2.2.03.08c.vii 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. Nuclear Construction and Startup Procedure (NCSP) 3-24, Field Surveying
2. VEGP 3&4 Updated Final Safety Analysis Report (UFSAR), Revision 11.2, Subsection 6.3.2.2.7 IRWST and Containment Recirculation Screens
3. SV4-PXS-M6K-900192, Rev. 0, PXS ITAAC 192 Cont. Recirc. Screens ITAAC Inspections/Field Measurements
4. 2.2.03.08c.vii-U4-CP-Rev0, ITAAC Completion Package