ML21316A057

NRC Initial Test Program and Operational Programs Integrated Inspection Reports 05200025/2021005, 05200026/2021005

ML21316A057
Person / Time
Site:	Vogtle
Issue date:	11/12/2021
From:	Bradley Davis NRC/RGN-II
To:	Yox M Southern Nuclear Operating Co
References
IR 2021005
Download: ML21316A057 (51)

Text

November 12, 2021 Mr. Michael Yox Regulatory Affairs Director Southern Nuclear Operating Company 7825 River Road, BIN 63031 Waynesboro, GA 30830

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 - NRC INITIAL TESTPROGRAM AND OPERATIONAL PROGRAMS INTEGRATED INSPECTION REPORTS 05200025/2021005, 05200026/2021005

Dear Mr. Yox:

On September 30, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Vogtle Electric Generating Plant, Units 3 and 4. The enclosed inspection report documents the inspection results, which the inspectors discussed on October 18, 2021 with Mr.

Glen Chick, Vogtle 3 & 4 Executive Vice President, and other licensee and contractor staff members.

The inspection examined a sample of construction activities conducted under your Combined License (COL) as it relates to safety and compliance with the Commissions rules and regulations and with the conditions of these documents. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

NRC inspectors documented five findings of very low safety significance (Green) in this report.

All of these findings involved a violation of NRC requirements. The NRC is treating these violations as noncited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system ADAMS. ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Should you have any questions concerning this letter, please contact us.

Sincerely,

/RA/

Bradley J. Davis, Chief Construction Inspection Branch 2 Division of Construction Oversight Docket Nos.: 5200025, 5200026 License Nos: NPF-91, NPF-92

Enclosure:

NRC Inspection Report (IR) 05200025/2021005, 05200026/2021005 w/attachment: Supplemental Information

cc w/ encls:

Resident Manager Resident Inspector Oglethorpe Power Corporation Vogtle Plant Units 3 & 4 Alvin W. Vogtle Nuclear Plant 8805 River Road 7821 River Road Waynesboro, GA 30830 Waynesboro, GA 30830 Mr. Barty Simonton Office of the Attorney General Team Leader 40 Capitol Square, SW Environmental Radiation Program Atlanta, GA 30334 Air Protection Branch Environmental Protection Division Southern Nuclear Operating Company 4244 International Parkway, Suite 120 Document Control Coordinator Atlanta, GA 30354-3906 3535 Colonnade Parkway Birmingham, AL 35243 Brian H. Whitley Regulatory Affairs Director Anne F. Appleby Southern Nuclear Operating Company Olgethorpe Power Corporation 3535 Colonnade Parkway, BIN N-226-EC 2100 East Exchange Place Birmingham, AL 35243 Tucker, GA 30084 Mr. Michael Yox County Commissioner Regulatory Affairs Director Office of the County Commissioner Southern Nuclear Operating Company Burke County Commission 7825 River Road, BIN 63031 Waynesboro, GA 30830 Waynesboro, GA 30830 Mr. Wayne Guilfoyle Commissioner District 8 Augusta-Richmond County Commission 4940 Windsor Spring Rd Hephzibah, GA 30815 Gwendolyn Jackson Burke County Library 130 Highway 24 South Waynesboro, GA 30830 Mr. Reece McAlister Executive Secretary Georgia Public Service Commission Atlanta, GA 30334

Email aagibson@southernco.com (Amanda Gibson) acchambe@southernco.com (Amy Chamberlian) bhwhitley@southernco.com (Brian Whitley)

Bill.Jacobs@gdsassociates.com (Bill Jacobs) corletmm@westinghouse.com (Michael M. Corletti) crpierce@southernco.com (C.R. Pierce) dahjones@southernco.com (David Jones) david.hinds@ge.com (David Hinds) david.lewis@pillsburylaw.com (David Lewis) dlfulton@southernco.com (Dale Fulton) ed.burns@earthlink.net (Ed Burns) edavis@pegasusgroup.us (Ed David)

G2NDRMDC@southernco.com (SNC Document Control)

George.Taylor@opc.com (George Taylor) harperzs@westinghouse.com (Zachary S. Harper) james1.beard@ge.com (James Beard)

JHaswell@southernco.com (Jeremiah Haswell) jim@ncwarn.org (Jim Warren)

John.Bozga@nrc.gov (John Bozga)

Joseph_Hegner@dom.com (Joseph Hegner) karlg@att.net (Karl Gross) kmstacy@southernco.com (Kara Stacy) kroberts@southernco.com (Kelli Roberts)

KSutton@morganlewis.com (Kathryn M. Sutton) kwaugh@impact-net.org (Kenneth O. Waugh) markus.popa@hq.doe.gov (Markus Popa) mdmeier@southernco.com (Mike Meier) media@nei.org (Scott Peterson)

Melissa.Smith@Hq.Doe.Gov (Melissa Smith) mike.price@opc.com (M.W. Price)

MKWASHIN@southernco.com (MKWashington) mphumphr@southernco.com (Mark Humphrey)

MSF@nei.org (Marvin Fertel) nirsnet@nirs.org (Michael Mariotte)

Nuclaw@mindspring.com (Robert Temple)

Paul@beyondnuclear.org (Paul Gunter) pbessette@morganlewis.com (Paul Bessette) ppsena@southernco.com (Peter Sena,III) r.joshi15@comcast.net (Ravi Joshi) rwink@ameren.com (Roger Wink) sabinski@suddenlink.net (Steve A. Bennett) sjackson@meagpower.org (Steven Jackson) sjones@psc.state.ga.us (Shemetha Jones) skauffman@mpr.com (Storm Kauffman) slieghty@southernco.com (Steve Leighty) sroetger@psc.state.ga.us (Steve Roetger) syagee@southernco.com (Stephanie Agee)

TomClements329@cs.com (Tom Clements)

Vanessa.quinn@dhs.gov (Vanessa Quinn) wayne.marquino@gmail.com (Wayne Marquino)

William.Birge@hq.doe.gov (William Birge)

X2edgran@southernco.com (Eddie R. Grant) x2gabeck@southernco.com (Gary Becker)

X2hagge@southern.com (Neil Haggerty)

X2wwill@southernco.com (Daniel Williamson)

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 - NRC INITIAL TEST PROGRAM AND OPERATIONAL PROGRAMS INTEGRATED INSPECTION REPORTS 05200025/2021004, 05200026/2021004 DATED: November 12, 2021 DISTRIBUTION:

A. Veil, NRR G. Bowman, NRR V. Hall, NRR M. Webb, NRR T. Fredette, NRR G. Armstrong, NRR E. Duncan, NRR L. Dudes, RII J. Munday, RII M. Bailey, RII O. Lopez-Santiago, RII B. Davis, RII R2DCO PUBLIC Accession No.ML ____________________________________

OFFICE NAME C. Even J. Eargle J. Parent B. Davis DATE 11/11/2021 11/11/021 11/11/2021 11/12/2021 OFFICIAL RECORD COPY

U.S. NUCLEAR REGULATORY COMMISSION Region II Docket Numbers: 5200025 5200026 License Numbers: NPF-91 NPF-92 Report Numbers: 05200025/2021005 05200026/2021005 Licensee: Southern Nuclear Operating Company, Inc.

Facility: Vogtle Unit 3 & 4 Combined License Location: Waynesboro, GA Inspection Dates: July 1, 2021 through September 30, 2021 Inspectors: J. Eargle, Senior Resident Inspector - Testing, Division of Construction Oversight (DCO)

S. Downey, Senior Reactor Inspector, Division of Reactor Safety (DRS)

J. Dymek, Reactor Inspector, DRS C. Even, Senior Construction Inspector, DCO T. Fanelli, Senior Reactor Inspector, DRS M. Greenleaf, Technical Assistant, Division of Fuel Facility Inspection B. Griman, Construction Inspector, DCO N. Hansing, Mechanical Engineer, Nuclear Reactor Regulation (NRR)

B. Kemker, Senior Resident Inspector, DCO M. Magyar, Construction Inspector, DRS J. Montgomery, Senior Reactor Inspector, DRS T. Morrissey, Senior Construction Inspector, DCO J. Parent, Resident Inspector, DCO R. Patel, Senior Construction Inspector, DCO R. Patterson, Senior Reactor Inspector, DRS M. Riley, Senior Construction Inspector, DCO T. Scarbrough, Senior Mechanical Engineer, NRR Approved by: Bradley J. Davis, Chief Construction Inspection Branch 2 Division of Construction Oversight

SUMMARY

OF FINDINGS Inspection Report (IR) 05200025/2021005, 05200026/2021005; 07/01/2021 through 09/30/2021; Vogtle Unit 3 & 4 COL, initial test program and operational programs integrated inspection report.

This report covers a three-month period of announced inspection Inspections, Tests, Analysis, and Inspection Criteria (ITAAC), preoperational test program, startup test program, and operational program inspections by resident and regional inspectors. Five findings were determined to be of very low safety significance (Green) by the inspectors. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 2519, Construction Significance Determination Process. Cross-cutting aspects are determined using IMC 0613, Appendix F, Construction Cross-Cutting Areas and Aspects.

All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy and the temporary enforcement guidance outlined in enforcement guidance memorandum 11-006. The NRCs program for overseeing the safe construction of commercial nuclear power reactors is described in IMC 2506, Construction Reactor Oversight Process General Guidance and Basis Document.

A. NRC-Identified and Self Revealed Findings (Green) A self-revealed construction finding of very low safety significance with an associated noncited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to accomplish activities in procedure 3-IDSC-SOP-002, Class 1E AC System

- Division C, Version 0 in accordance with procedure NMP-AP-003, Procedure and Work Instruction Use and Adherence, Version 6.1. The licensee entered this issue into its corrective action program (CAP) as condition report (CR) 50103788, CR 50103789, and CR 50103907.

The performance deficiency was determined to be of more than minor significance and thus a finding because it represented a substantive failure to implement an adequate program, process, procedure, or quality oversight function. The inspectors determined the finding was of very low safety significance (Green) because the event impaired only one train of the IDS. The inspectors determined the finding had a cross-cutting aspect of H.2, Field Presence, in the area of Human Performance. Specifically, senior managers did not ensure that operators performing 3-IDS-SOP-002 had adequate supervisory and management oversight of work activities which led to plant equipment being damaged. [H.2] (Section 1P01)

(Green) The NRC inspectors identified a performance deficiency and construction finding of very low safety significance and an associated Severity Level (SL) IV non-cited violation (NCV) of 10 CFR 50.59, Changes, tests, and experiments, for the licensees failure to perform a written safety evaluation prior to implementing a change to remove ASME QME-1-2007 commitments from the UFSAR in accordance with procedure ND-LI-VNP-002. The licensee entered this issue into its corrective action program as CR 50098695.

The inspectors determined that the failure to follow procedure ND-LI-VNP-002 to perform a 10 CFR 50.59 evaluation for the removal of ASME QME-1-2007 commitments from the design bases was a violation of 10 CFR 50.59(d)(1), and a performance deficiency. Per the guidance in IMC 0613, Appendix E, Examples of Minor Construction Issues, the performance deficiency

was determined to be more than minor, and thus a finding, because it represented a substantive failure to establish or implement an adequate program, process, procedure, or quality oversight function. Specifically, the removal of QME-1-2007 for dynamic restraints constituted a more than minimal increase in the likelihood of a malfunction of a structure, system, or component important to safety and there was a reasonable likelihood that the change would have required Commission review and approval prior to implementation in accordance with 10 CFR 50.59(c)(2)(ii). The inspectors determined the finding had a cross-cutting aspect of H.3, Change Management in the area of Human Performance. Specifically, the licensee failed to ensure the process for reviewing the change to remove QME-1-2007 for dynamic restraints was systematic in that all aspects of QME-1-2007 were evaluated for their impact on dynamic restraints. [H.3]

(Section 3P01)

(Green) NRC inspectors identified a performance deficiency and ITAAC finding of very low safety significance and an NCV of 10 CFR 50.49(e)(5), Environmental qualification of electric equipment important to safety for nuclear power plants, for the licensees failure to establish the qualified lives of all nonmetallic components necessary for the completion of the safety function of the squib and main steam isolation (MSI) valves in accordance with procedure APP-GW-G1-002, "AP1000 Equipment Qualification Methodology." The licensee entered this issue into its CAP as CR 50096894.

The performance deficiency was determined to be more than minor, and thus a finding, because it is material to the acceptance criteria of the ITAAC, and the performance deficiency prevented the licensee from meeting the ITAAC Design Commitment. The inspectors determined this finding was of very low safety significance because the finding did not impair a design function of the valves. The inspectors determined no cross-cutting aspect applied because the performance deficiency did not reflect current licensee performance. (Section 3P01)

(Green) NRC inspectors identified an ITAAC finding and associated NCV of 10 CFR 50.49.e(5) for the failure to establish the qualified life of the Containment High Range Monitor (CHRM) door gasket in accordance with Institute of Electrical and Electronics Engineers Standard (IEEE) 323-1974. The licensee entered this issue into its CAP as CR 50096896 and CAP-IR-2021-6870 and is working on corrective actions to address the issue.

The performance deficiency was determined to be more than minor, and thus a finding, because it represented an adverse condition that rendered the quality of a system, structure, or component (SSC), unacceptable or indeterminate, and required substantive corrective action. The inspectors determined that the finding was of very low safety significance because the finding was not associated with a system or structure in risk importance table. The inspectors determined the finding was associated with the cross-cutting aspect of Consistent Process in the area of Human Performance. Specifically, the licensee failed to use a consistent, systematic approach to make decisions whether the AP1000 harsh environment included wetted environments. [H.13] (Section 3P01)

(Green) NRC identified an ITAAC finding and associated violation of 10 CFR 50, Appendix B, Criterion III, for the failure to seismically or environmentally qualify the Electrical Penetration Assemblies (EPAs) in accordance with IEEE 317, Section 6.2.10. Specifically, the licensee failed to test or analyze the as installed assembly including cable configurations, terminations, termination box, raceways, and their respective interactions. The licensee entered this issue into its CAP as CR 50104476 and is working on corrective actions to address the issue

The performance deficiency was determined to be more than minor, and thus a finding, because it represented an adverse condition that rendered the quality of an SSC unacceptable or indeterminate and required substantive corrective action. The inspectors determined the finding was of very low safety significance because the licensee was able to show through additional analysis that the finding did not impact a design function of the EPAs. The inspectors determined the finding was associated with the cross-cutting aspect of Change Management in the area of Human Performance. Specifically, the licensee leadership failed to use a systematic process for evaluating and implementing changes to as installed SSCs compared to configurations that were environmentally qualified so that nuclear safety remained the overriding priority. [H.3] (Section 3P01)

B. Licensee-Identified Violations None

REPORT DETAILS Summary of Plant Construction Status During this report period for Unit 3, the licensee completed various activities to satisfy aspects of the Vogtle Unit 3 operational programs and initial test program. The licensee completed hot functional testing activities which included testing the reactor coolant system, residual heat removal system, passive core cooling system, main steam system, etc. at elevated temperatures and pressures. The licensee performed post hot functional testing inspection of the reactor internals to ensure that there was no observable damage or loose parts. The licensee performed preoperational and component tests of various SSCs and their control systems, e.g. Class 1E DC and UPS system (IDS), protection and monitoring system (PMS),

and plant control system. Additionally, the licensee performed testing of air operated valves and motor operated valves; and performed response time testing of permanently installed plant instrumentation.

During this report period for Unit 4, the licensee continued with integrated flush activities by flushing portions of chemical and volume control system spent fuel pool system (SFS), and RNS. The licensee completed the initial pump runs for the component cooling water system and service water system, and the initial compressor run for the compressed and instrument air system. In addition to continuing with the installation of plant SSCs, the licensee is making preps for the upcoming secondary hydrostatic test.

1. CONSTRUCTION REACTOR SAFETY Cornerstones: Design/Engineering, Procurement/Fabrication, Construction/Installation, Inspection/Testing IMC 2503, Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) - Related Work Inspections 1A01 (Unit 4) ITAAC Number C.2.6.09.08a (668) / Family 17A

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number C.2.6.09.08a (668). The inspectors used the following NRC inspection procedure (IP)/sections to perform this inspection:

• 65001.17 - Inspection of ITAAC-related Security Structure, Systems, and Components The inspectors performed an inspection to determine if the openings in Vogtle Unit 4 vital area barriers for heating, ventilation, and cooling system vents were secured to prevent exploitation of the openings to satisfy the ITAAC and 10 CFR 73.55(e)(4).

The inspectors reviewed the design specifications and associated drawings to identify designated heating, ventilation, and cooling system openings through vital area barriers

and the way they will be secured and monitored. The inspectors examined the physical installation of one heating, ventilation, and cooling system opening (SV4-VAS-AS-04) to the vital area during this inspection period. The inspectors performed direct observation inspection of the opening to determine if it was secured in a manner that would delay or prevent exploitation. Specifically, the inspectors directly inspected the barrier, locking mechanisms, welds, and bolts associated with the opening.

b. Findings No findings were identified.

IMC 2504, Construction Inspection Program - Inspection of Construction and Operational Programs 1P01 Construction Quality Assurance (QA) Criterion 16

• 35007-A16.04 - Inspection Requirements and Guidance

• 35007-A16.04.02 - Inspection of QA Program Implementation

a. Inspection Scope The inspectors conducted daily reviews of issues entered into the licensees CAP to assess issues that might warrant additional follow-up inspection, to assess repetitive or long-term issues, to assess adverse performance trends, and to ensure the CAP appropriately included regulatory required non-safety related SSCs. The inspectors periodically attended the licensees CAP review meetings, held discussions with licensee and contractor personnel, and performed reviews of CAP activities during the conduct of other baseline inspection procedures. The inspectors reviewed conditions entered into the licensees CAP to determine whether the issues were classified in accordance with the licensees quality assurance (QA) program and CAP implementing procedures. The inspectors reviewed corrective actions associated with conditions entered into the CAP to determine whether appropriate actions to correct the issues were identified and implemented effectively, including immediate or short-term corrective actions, in accordance with the applicable QA program requirements and 10 CFR 50, Appendix B, Criterion XVI. Additionally, the inspectors reviewed the corrective actions taken to determine whether they were commensurate with the significance of the associated conditions in accordance with the licensees CAP implementing procedures. The inspectors completed reviews of CAP entry logs to verify if issues from all aspects of the project, including equipment, human performance, and program issues, were being identified by the licensee and associated contractors at an appropriate threshold and entered into the CAP as required by licensees CAP implementing procedures. The inspectors performed a focused review on the following condition reports:

• CR 50103788;

• CR 50103789; and

• CR 50103907.

b. Findings Introduction A self-revealed construction finding of very low safety significance (Green) with an associated noncited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to accomplish activities in procedure 3-IDSC-SOP-002, Class 1E AC System - Division C, Version 0 in accordance with procedure NMP-AP-003, Procedure and Work Instruction Use and Adherence, Version 6.1.

Description On August 12, 2021, operators were removing IDSC-DU-1 DIV C, 120 VAC Inverter 1, and 3-IDSC-EA-1 DIV C, 120 VAC Distribution Panel 1, from service per procedure 3-IDSC-SOP-002. This procedure was designated as a continuous use procedure.

Southern Nuclear Company (SNC) procedure NMP-AP-003, Section 4.3, specifies, in part, that for a continuous use procedure, operators perform all steps in sequence, complete each step before moving to the next step, and sign off each step after completing it and before moving to the next step.

During the performance of procedure 3-IDSC-SOP-002, operators performed Attachment 4, Section 4.2, Step d, to discharge the inverters capacitor, before completing Step c, to open the inverters DC input breaker. As a result of performing these steps out of sequence, the internal resistors for the inverter failed and require replacement. The licensee entered this issue into its CAP as CR 50103788, CR 50103789, and CR 50103907.

Analysis The inspectors determined the failure to accomplish activities in procedure 3-IDSC-SOP-002 in accordance with procedure NMP-AP-003 was contrary to 10 CFR 50, Appendix B, Criterion V and was a performance deficiency. The performance deficiency was determined to be of more than minor significance and thus a finding because it represented a substantive failure to implement an adequate program, process, procedure, or quality oversight function. Specifically, the licensee failed to perform all steps in sequence, complete each step before moving to the next step, and sign off each step after completing it and before moving to the next step. This caused damage to the inverter due to the internal resistors failing.

The inspectors concluded the finding was associated with the Construction/Installation Cornerstone and assessed the finding in accordance with IMC 2519, Construction Significance Determination Process, Appendix A, AP 1000 Construction Significance Determination Process, Section 4. The inspectors determined the finding was associated with the IDS and was of very low safety significance because the event impaired only one train of the IDS.

The performance deficiency did not impact an ITAAC, thus it was determined to be a construction finding.

In accordance with IMC 0613 Appendix F, Construction Cross-Cutting Areas and Aspects, the inspectors determined the finding had a cross-cutting aspect of H.2, Field Presence, in the area of Human Performance. Specifically, senior managers did not ensure that operators performing 3-IDS-SOP-002 had adequate supervisory and management oversight of work activities which led to plant equipment being damaged.

Enforcement 10 CFR 50, Appendix B, Criterion V, states in part, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Contrary to the above, on August 12, 2021, the licensee failed to accomplish activities in procedure 3-IDSC-SOP-002 in accordance with procedure NMP-AP-003.

Specifically, when performing 3-IDSC-SOP-002, the licensee failed to perform all steps in sequence, complete each step before moving to the next step, and sign off each step after completing it and before moving to the next step. This finding did not present an immediate safety concern because the issue only affected one of the four divisions, the plant was not operating, and the reactor vessel did not have fuel in it. The inverter was immediately secured and placed the affected division in a safe condition, and further corrective actions that are being executed and tracked by the licensee include troubleshooting and repairs to the inverter's resistors and personnel training. The licensee entered this issue into its CAP as CR 50103788, CR 50103789, and CR 50103907.

Because this violation was not repetitive or willful, was of very low safety significance (Green), and was entered into the licensees corrective action program, this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy (NCV 05200025/2021005-01, Failure to Follow Procedure 3-IDSC-SOP-001).

3. OPERATIONAL READINESS Cornerstones: Operational Programs IMC 2503, Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) - Related Work Inspections 3T01 (Unit 3) ITAAC Number 2.1.02.08b (30) / Family 06D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.1.02.08b (30). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.03-Test Results Review

The inspectors used appropriate portions of the IP to review the results of the following procedure to determine if the reactor coolant pumps (RCPs) have a rotating inertia to provide reactor coolant system (RCS) flow coastdown on loss of power to the pumps.

Specifically, the test results were reviewed to verify if they satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC.

• 3-RSC-ITPP-506, Reactor Coolant Pump and Reactor Coolant Flow Precore Hot Functional, Version (Ver.) 3

b. Findings No findings were identified.

3T02 (Unit 3) ITAAC Number 2.1.02.09a (41) / Family 14D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.1.02.09a (41). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.03-Test Results Review The inspectors used appropriate portions of the IP to review the results of the following procedure to determine if the calculated post-fuel load RCS flow rate was greater than or equal to 301,670 gallons per minute. Specifically, the test results were reviewed to verify if they satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC:

• 3-RCS-ITPP-506, Reactor Coolant Pump and Reactor Coolant Flow Precore Hot Functional, Ver. 3

b. Findings No findings were identified.

3T03 (Unit 3) ITAAC Number 2.1.02.11a.ii (47) / Family 10C

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.1.02.11a.ii (47). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.03-Test Results Review

The inspectors used the appropriate portions of the IP to review the results of the following procedure to determine if the remotely operated valves identified as having PMS control performed their active function after receiving a signal from PMS, and that these valves opened within the required times after receipt of an actuation signal.

Specifically, the test results were reviewed to verify if they satisfied the applicable quality and technical requirements of the UFSAR and ITAAC.

• B-GEN-ITPCI-039, PMS CIM Component Test Procedure, Ver. 3

b. Findings No findings were identified.

3T04 (Unit 3) ITAAC Number 2.1.02.12a.iii (55) / Family 07D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.1.02.12a.iii (55). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.03-Test Results Review The inspectors used appropriate portions of the IP to review the results of the following procedure to determine if the automatic depressurization system stages 1, 2, and 3 motor operated valves changed position under preoperational test conditions.

Specifically, the test results were reviewed to verify if they satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC.

• 3-PXS-ITPP-505, "ADS Stages 1-3 Dynamic Test," Revision (Rev.) 1.1

b. Findings No findings were identified.

3T05 (Unit 3) ITAAC Number 2.1.03.07.i (78) / Family 05D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.1.03.07.i (78). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.02-Test Witnessing The inspectors used appropriate portions of the IP to observe the licensees performance of the following procedure used to test if the reactor internals would withstand the effects of flow induced vibration. Specifically, the tests were observed to Security Related Information - Official Use Only

Security Related Information - Official Use Only verify if they satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC.

• SV3-CVAP-T2R-100, Vogtle Unit 3 Pre-Hot Functional Test Inspection of Reactor Vessel Internals, Rev. 3.0

b. Findings No findings were identified.

3T06 (Unit 3) ITAAC Number 2.2.01.11a.iii (116) / Family 07D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.2.01.11a.iii (116). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.02-Test Witnessing

• 65001.D-02.03-Test Results Review The inspectors used appropriate portions of the IP to observe the licensees performance of the following procedure used to test if the motor-operated valve performed its active safety-related function to change position under preoperational conditions. Specifically, the tests were observed to verify if they satisfied the applicable quality and technical requirements of the Updated Final Safety Analysis Report (UFSAR) and the ITAAC.

• 3-CCS-ITPP-501, Component Cooling Water System Preoperational Test Procedure, Rev. 2.0 The inspectors used appropriate portions of the IP to review the results of the following procedures to determine if the motor-operated valves performed their active safety-related function to change position. Specifically, the tests were observed to verify if they satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC.

• G-GEN-ITPCM-001, Limitorque SMB/SB Motor Operated Valve Component Testing, Ver. 3.0

• 3-SFS-ITPP-502, Spent Fuel Pool Cooling System Flow Path Preoperational Test Procedure, Ver. 3.1

b. Findings No findings were identified.

3T07 (Unit 3) ITAAC Number 2.2.01.11b (118) / Family 07D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.2.01.11b (118). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.03-Test Results Review The inspectors used appropriate portions of the IP to review the results of the following procedures to determine if after loss of motive power, the remotely operated valves assumed the indicated loss of motive power position. Specifically, the test results were reviewed to verify if they satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC.

• B-GEN-ITPCM-017, Air-Operated Valve Test, Ver. 6.1

• B-GEN-ITPCM-001, Limitorque SMB/SB Motor Operated Valve Component Testing, Ver. 4

b. Findings No findings were identified.

3T08 (Unit 3) ITAAC Number 2.2.03.08b.01 (175) / Family 06D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.2.03.08b.01 (175). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.02-Test Witnessing

• 65001.D-02.03-Test Results Review The inspectors used the appropriate portions of the IP to observe the licensee's performance of the following procedure used to verify if the passive residual heat removal heat exchanger (PRHR HX) heat transfer rate exceeded the heat transfer rate predicted for the as-tested initial conditions. Specifically, the test was observed to verify if the test satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC.

• 3-PXS-ITPP-504, Passive Core Cooling System Hot Functional Test, Rev. 2.1 The inspectors also used appropriate portions of the IP to review the test results that verified if the PRHR HX heat transfer rate exceeded the heat transfer rate predicted for the as-tested initial conditions. Specifically, the test results were reviewed to determine whether they contained sufficient information to meet the requirements of the UFSAR and ITAAC acceptance criteria.

• 3-PXS-ITPP-504, Passive Core Cooling System Hot Functional Test, Rev. 2.1

b. Findings No findings were identified.

3T09 (Unit 3) ITAAC Number 2.6.03.04c (603) / Family 08D

a. Inspection Scope The inspectors performed a direct inspection of construction activities associated with ITAAC Number 2.6.03.04c (603). The inspectors used the following NRC IPs/sections to perform this inspection:

• 65001.D-02.02-Test Witnessing The inspectors used appropriate portions of the IP to observe the licensees performance of the following procedures used to test if the battery terminal voltage was greater than or equal to 210 Volts after a period of no less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with an equivalent load that equals or exceeds the battery bank design duty cycle capacity.

Specifically, the tests were observed to verify if they satisfied the applicable quality and technical requirements of the UFSAR and the ITAAC.

• B-GEN-ITPCE-008, "Class 1E and Non-Class 1E Battery testing,"(Division B 125 VDC 24hr Battery 1B, and Division C 250 VDC 24hr and 72hr service tests), Ver. 7

b. Findings No findings were identified.

IMC 2504, Construction Inspection Program - Inspection of Construction and Operational Programs 3P01 Environmental Qualification

• 51080-02.03 - Inspection Tasks

• 51080-App A - Checklist for Review of Licensee Electrical Environmental Qualification (EQ) Documentation Files

• 51080-App C - EQ of Nonmetallic Parts for Pumps, Valves, and Dynamic Restraints

a. Inspection Scope The inspectors reviewed the EQ program to verify if the program was established, implemented, and documented in accordance 10 CFR 50.49. The inspectors reviewed program procedures to determine if the licensee had implemented a program to generate, maintain, and distribute the list of electrical and mechanical equipment

requiring EQ. The inspectors reviewed procedures for EQ of nonmetallic parts for pumps, valves, and dynamic restraints to verify if those components were captured in the EQ program in accordance with American Society of Mechanical Engineers (ASME)

QME-1-2007. The inspectors sampled EQDPs for low voltage cable splices, junction boxes, and EPAs to verify if the components required for EQ were tested or analyzed in accordance with 10 CFR 50.49.

b. Findings NCV 05200025/2021005-02, Failure to Complete 10 CFR 50.59 Evaluation Prior to Altering Dynamic Restraint Qualification Requirements Introduction The NRC inspectors identified a performance deficiency and construction finding of very low safety significance (Green) and an associated Severity Level (SL) IV non-cited violation (NCV) of 10 CFR 50.59, Changes, tests, and experiments, for the licensees failure to perform a written safety evaluation prior to implementing a change to remove ASME QME-1-2007 commitments from the UFSAR in accordance with procedure ND-LI-VNP-002.

Description For initial licensing, the NRC staff relied upon the content of the UFSAR that specified that dynamic restraints would be qualified in accordance with ASME QME-1 2007. After the initial licensing review, the licensee initiated LDCR-2016-107 to remove ASME QME-1-2007 from the UFSAR and thus the licensing basis for dynamic restraints under 10 CFR 50.59. The licensee incorrectly determined it would not be an adverse change to the licensing basis. Changing the qualification basis from ASME QME-1-2007 to ASME BPV Section III removed qualified life requirements for the non-metallic parts of the dynamic restraints.

The licensee compared ASME QME-1-2007, Section QDR, Qualification of Dynamic Restraints, to ASME Boiler and Pressure Vessel Code (BPV),Section III, Subsection NF, (1998 with 2002). The ASME BPV,Section III covered material compatibility, structural integrity, etc. but did not establish qualified life requirements for the non-metallic parts of dynamic restraints. The licensee failed to consider the lack of requirements for non-metallic parts in ASME BPV,Section III that are in ASME QME 2007.

In addition, Section QR in ASME-QME-1-2007 described acceptable methods for qualifying non-metallic parts in pumps, valves, and dynamic restraints without conflict with Section QDR. Following the change, neither the licensees FSAR nor ASME Boiler and Pressure Vessel Code Section III addressed qualification of non-metallic parts for dynamic restraints. The inspectors concluded that the removal of qualification requirements for dynamic restraints was an adverse change to the facility as described in the UFSAR as it decreased the reliability of dynamic restraint design function over their designated lives.

Based on NEI 96-07, Section 4.3.2, the inspectors concluded that the removal of qualification requirements from dynamic restraints was an adverse change and that a 10 CFR 50.59 safety evaluation should have been performed prior to implementing the change. The inspectors also concluded that the change constituted a more than minimal increase in the likelihood of a malfunction of a structure, system, or component important to safety since the change failed to ensure that the licensee still met other acceptance criteria in ASME QME-1-2007, to which they were committed. As this change was a more than minimal increase, there was a reasonable likelihood that the change would have required Commission review and approval prior to implementation in accordance with 10 CFR 50.59(c)(2)(ii).

The inspectors reviewed licensee procedure ND-LI-VNP-002 and determined that the licensee did not follow its procedure. The procedure made the appropriate references to NEI 96-07 and provided sufficient explanation for the licensee to determine that the change was adverse and that a written evaluation should have been performed. The procedure also had sufficient explanation for the licensee to determine that the change resulted in a more than minimal increase in the likelihood of a malfunction of a structure, system, or component, and that there was a reasonable likelihood that the change required a license amendment. The licensee entered this issue into its corrective action program as CR 50098695.

Analysis The inspectors determined that the failure to follow procedure ND-LI-VNP-002 to perform a 10 CFR 50.59 evaluation for the removal of ASME QME-1-2007 commitments from the design bases was a violation of 10 CFR 50.59(d)(1), and a performance deficiency. Per the guidance in IMC 0613, Appendix E, Examples of Minor Construction Issues, the performance deficiency was determined to be more than minor, and thus a finding, because it represented a substantive failure to establish or implement an adequate program, process, procedure, or quality oversight function.

Specifically, the removal of QME-1-2007 for dynamic restraints constituted a more than minimal increase in the likelihood of a malfunction of a structure, system, or component important to safety and there was a reasonable likelihood that the change would have required Commission review and approval prior to implementation in accordance with 10 CFR 50.59(c)(2)(ii).

This finding was a construction finding because the performance deficiency was not material to the acceptance criteria of an ITAAC.

The inspectors determined the finding was associated with the Inspection/Testing Cornerstone and assessed the finding in accordance with IMC 2519, Construction Significance Determination Process, Appendix A, AP1000 Construction Significance Determination Process, Section 4. The inspectors determined this finding was of very low safety significance because the finding did not impair a design function.

In accordance with IMC 0613 Appendix F, Construction Cross-Cutting Areas and Aspects, the inspectors determined the finding had a cross-cutting aspect of H.3, Change Management in the area of Human Performance. Specifically, the licensee failed to ensure the process for reviewing the change to remove QME-1-2007 for

dynamic restraints was systematic in that all aspects of QME-1-2007 were evaluated for their impact on dynamic restraints.

In addition, the Construction Reactor Oversight Process significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.

The inspectors determined that the failure to perform a 10 CFR 50.59 evaluation for the removal of ASME QME-1-2007 from the design bases was a violation of 10 CFR 50.59(d)(1). In accordance with Section 6.5 of the NRC Enforcement Policy, the inspectors determined this was a Severity Level IV violation because the conditions of the performance deficiency were evaluated as having very low safety significance.

Enforcement 10 CFR 52.98 (c)(2) states, in part, that changes that are not within the scope of the referenced design certification rule are subject to the applicable change processes in 10 CFR part 50, unless they also involve changes to or noncompliance with information within the scope of the referenced design certification rule 10 CFR 50.59(d)(1), Changes, tests, and experiments, requires, in part, that licensees shall maintain records of changes to the facilitymade pursuant to paragraph (c) of this section. These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c)(2) of this section.

Contrary to the above, since April 2020, the licensee failed to include a written evaluation prior to making a change to the facility as described in the UFSAR. Specifically, the licensee failed to perform a 10 CFR 50.59 evaluation prior to implementing a change to remove ASME QME-1-2007 from the UFSAR.

The licensee entered this issue into its CAP as CR 50098695 and is working on corrective actions. This issue does not represent an immediate safety concern because there was no fuel in the reactor and the dynamic restraints were not required to perform their safety function at the time of the inspection. Because this violation was not repetitive or willful, and was entered into the licensees CAP, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy (NCV 05200025/2021005-02, Failure to Obtain a License Amendment Prior to Altering Dynamic Restraint Qualification Requirements).

NCV 05200025/2021005-03, Failure to Demonstrate Qualification of Valve Nonmetallic Parts Introduction The NRC inspectors identified a performance deficiency and ITAAC finding of very low safety significance (Green) and an NCV of 10 CFR 50.49(e)(5), Environmental qualification of electric equipment important to safety for nuclear power plants, for the

licensees failure to establish the qualified lives of all nonmetallic components necessary to the completion of the safety function of the squib and main steam isolation (MSI) valves in accordance with procedure APP-GW-G1-002, "AP1000 Equipment Qualification Methodology." Specifically, a qualified life for the Viton nonmetallic components used in the squib and MSI valves was not established.

Description During the review of the qualification reports for the squib and MSI valves, the inspectors determined that the non-metallic parts, specifically Viton seals, were tested without establishing that they had been subjected to artificial aging sufficient to envelop the thermal degradation expected over the installed life of the elastomer. Specifically, the MSI valve Viton elastomers were not thermally aged at all, while the Viton gaskets and O-rings in the squib valve assembly were only thermally aged to an equivalent period of approximately 50% of the desired qualified life.

Prior to testing the valves, the licensee made the determination that the Viton materials were insensitive to the deleterious effects of thermal aging based on the materials long expected life. This determination was not in accordance with licensee procedure APP-GW-G1-002, Sections 3.3.2, Aging, and 5.2.2, Thermal Aging, which determined that the aging mechanism for the non-metallic parts met the criteria as a significant aging mechanism and a conservative activation energy should have been applied in the selection and use of thermal aging parameters for test and calculations. Without aging or determining the activation energy of the Viton material, a qualified life wasnt determined.

In addition, the failure to artificially age the squib and MSI valves non-metallic parts failed to account for synergistic effects that are possible when multiple stress environments are applied simultaneously, such as aging, radiation, humidity, vibration, etc., in accordance with Section 3.3.2, Aging, in licensee procedure APP-GW-G1-002.

By failing to demonstrate that aging was not a significant aging mechanism and account for synergistic effects, the licensee was required to define a qualified life for the Viton O-rings or artificially age them sufficiently to their designated end-of-installed life condition. The licensee entered this issue into its corrective action program as CR 50096894.

Analysis The inspectors determined that the failure to follow procedure APP-GW-G1-002 to establish the qualified lives of all nonmetallic components necessary to the completion of the safety function of the squib and MSI valves was contrary to 10 CFR 50.49(e)(5) and was a performance deficiency. Per the guidance in IMC 0613, Appendix E, Examples of Minor Construction Issues, the performance deficiency was determined to be more than minor, and thus a finding, because it is material to the acceptance criteria of the ITAAC, and the performance deficiency prevented the licensee from meeting the ITAAC Design Commitment. Specifically, the licensee failed to demonstrate that the valves would be able to perform their design basis function for the qualified lives.

This finding is an ITAAC finding because the performance deficiency if material to ITAAC 2.2.04.05a.i and 2.1.02.05a.i. Specifically, the acceptance criteria requires that the components listed in Tables 2.2.4-1 and 2.1.2-1 can withstand the environmental conditions that would exist before, during, and after a design basis accident without loss of safety function for the time required to perform the safety function. By not thermally aging the Viton nonmetallic components, the licensee was not able to demonstrate it could meet the ITAAC.

The inspectors determined that the finding was associated with the Inspection/Testing Cornerstone and assessed the finding in accordance with IMC 2519, Construction Significance Determination Process, Appendix A, AP1000 Construction Significance Determination Process, Section 4. The inspectors determined this finding was of very low safety significance because the finding did not impair a design function of the valves.

The inspectors determined no cross-cutting aspect applied because the performance deficiency did not reflect current licensee performance.

Enforcement 10 CFR 50.49(e)(5) requires, in part, that equipment qualified by test must be preconditioned by natural or artificial aging to its end-of-installed life condition.

Contrary to the above, since 2012, the licensee failed to precondition Viton nonmetallic parts to their end-of-installed life condition. Specifically, for the qualification by test of the squib and MSI valves, the licensee failed to age the Viton elastomers, gaskets, and O-rings to their end-of-installed life conditions.

The licensee entered this issue into its corrective action program as CR 50096894 and is working on corrective actions to address the issue. This issue does not represent an immediate safety concern because there was no fuel in the reactor and these components were not required to perform their safety function at the time of the inspection.

Because this violation was not repetitive or willful, and was entered into the licensees corrective action program, this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy (NCV 05200025/2021005-03, Failure to Demonstrate Qualification of Valve Nonmetallic Parts).

NCV 05200025/2021005-04, Failure to Demonstrate Qualification of Containment High Radiation Monitor Door Gasket Introduction NRC identified a Green ITAAC finding and associated NCV of 10 CFR 50.49.e(5) for the failure to establish the qualified life of the Containment High Range Monitor (CHRM) door gasket in accordance with IEEE 323-1974.

Description

On July, 19, 2021, the inspectors performed an inspection of APP-RMS-VBR-002, Equipment Qualification Data Package for the Radiation Monitoring System Containment High Range Monitor and Field Cable Connections at the EPA Feedthroughs for the AP1000 Plant, Rev 0. The CHRM was qualified in accordance with the licensing basis in standard IEEE 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, as endorsed by Regulatory Guide (RG) 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants revision 1. IEEE 323-1974, Section 5, Principles of Qualification, specified, in part, principles and procedures for demonstrating the qualification of Class 1E equipment include: (2) Assurance that any extrapolation or inference be justified by allowances for known potential failure modes and the mechanism leading to them, (6) Qualification of any interfaces associated with Class 1E equipment. Examples of interfaces include connection boxes, splices, terminal boards, electrical connections, grommets, gaskets, cables, conduits, enclosures, etc.

The EQ documents specified that three different configurations of CHRM interfaces were tested for qualification. One configuration was labeled Equipment Under Test (EUT)1 that represented the installed interface configuration. EUT1 protected the CHRM soft components (connectors and cabling) with a metal connection box and rigid conduits and the various seals including the sensor door gasket. For the EUT2 and EUT3 configurations, the soft components were exposed to the LOCA environment, except for a piece of sheet metal to test the indirect effects of spray and steam on the components. The indirect effects of spray and steam significantly degraded the EUT2 and EUT3 interfaces causing failures of the CHRM functionality due to splitting and blistering in the soft components. Where EUT1 soft components were minimally exposed, due to the test setup, the spray and steam caused localized blistering and splitting in short sections of cabling. The inspectors noted that a qualified life of the door gasket interface even though it was present throughout the test preventing excessive moister from entering the connection box.

The evaluation of the failures and degradation was documented in report APP-RMS-J7C-009, AP1000 Radiation Monitoring System LOCA Test Engineering Evaluation for the Electrical Penetration Feedthroughs. The report Section 4, stated, in part, that there were no significant material differences in the connectors between the three configurations. Section 5.3, specified, that what caused the failures during the tests was the presence of spray rather than steam. Spray only needs to be postulated in the vicinity of the break [loss of coolant accident (LOCA) pipe break]. The inspectors noted that this statement regarding the spray only needs to be postulated in the vicinity of the break, inferred that spray is only associated with a LOCA pipe break and leaves out the containment spray systems. The spray simulation in the qualification test is for containment spray system in accordance with RG 1.89 regulatory position C.2.b. The AP1000 design is to condense and recirculate the LOCA atmosphere causing a showering effect similar to a containment spray system. The report, Section 5.3, indicated that it was the wetting of the components from Excessive Exposure (Spray rather than steam) and Excessive Exposure (Boric Acid and TSP rather than neutral),

that caused the EUT failures. Finally, the report, Section A.2, stated, in part, the EUT1 gasket is not a herm[e]tic seal. There is a path for communication between the outside environment and the environment inside the box. Mineral cloth wrapped around the cable connections was in good condition. Practically dry with only traces of soaking at the bottom.

The licensee justified not determining a qualified life for the door gasket based on the statements in Section A.2 above. The licensee inference was not justified because it failed to credit the door gasket preventing excessive moisture from entering the connection box during the test, and that spray was not a direct result of the LOCA break but rather containment spray similar to AP1000 condensate rain showering.

Not crediting the seals used during qualification did not meet the principles of qualification in IEEE 323-1974 Section 5. Without a gasket, the CHRM soft components could be exposed to the same degradation as the EUTs that failed the testing. The inspectors determined that the qualification for the CHRM, as tested, was in part based on the performance of the gasket, however, the qualification of the CHRM excluded the known potential failure modes (aging) of the gasket to determine the qualified life of the gasket as required by IEEE 323-1974. The licensee entered this issue into its CAP as CR 50096896 and CAP-IR-2021-6870 and is working on corrective actions to address the issue.

Analysis The failure to consider the known potential failure modes (aging) of the CHRM door gasket to determine a qualified life in accordance with IEEE 323-1974 was contrary to 10 CFR 50.49(e)5 and a performance deficiency. Per the guidance in IMC 0613, Appendix E, Examples of Minor Construction Issues, the performance deficiency was determined to be more than minor, and thus a finding, because it represented an adverse condition that rendered the quality of an SSC, unacceptable or indeterminate, and required substantive corrective action. Specifically, the failure to consider the known potential failure modes (aging) of the CHRM door gasket to determine a qualified life represented an adverse condition that rendered the quality of the CHRM unacceptable or indeterminate.

The inspectors determined the performance deficiency was material to the acceptance criteria of ITAAC index 823 (3.5.00.01.i). The applicable acceptance criterion of ITAAC 3.5.00.01.i requires, in part, that a report exists and concludes that Class 1E equipment identified in Table 3.5-1 as being located in a harsh environment can withstand the environmental conditions that would exist before, during, and following a design basis accident without loss of safety function for the time required to perform the safety function.

The inspectors determined that the CHRMs identified in Table 3.5-1 as PXS-RE (160, 161, 162, and 163) were not qualified to withstand the environmental conditions that would exist during and following a design basis accident without loss of safety function for the time required to perform the safety function.

The inspectors determined the finding was associated with the Construction/Installation cornerstone of the Construction Reactor Safety strategic performance area. The inspectors assessed the finding using IMC 2519, Appendix A, AP1000 Significance Determination Process, dated October 26, 2020, and determined that the finding was not associated with a security program determined, the finding was not associated with the failure to properly implement an IMC 2504 operational/construction program; and it

was not associated with a repetitive, NRC-identified omission of a program critical attribute.

The inspectors determined the finding was a performance deficiency with a significance very low safety significance, GREEN, because the finding was not associated with a system or structure in risk importance table.

The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of Consistent Process in the area of Human Performance, in accordance with IMC 0613, Appendix F, Construction Cross-Cutting Areas and Aspects, dated November 4, 2020. Specifically, the licensee failed to use a consistent, systematic approach to make decisions whether the AP1000 harsh environment included wetted environments. [H.13].

Enforcement 10 CFR Part 50.49.e(5), Aging, required, in part, that equipment qualified by test must be preconditioned by natural or artificial (accelerated) aging to its end-of-installed life condition. Consideration must be given to all significant types of degradation which can have an effect on the functional capability of the equipment.

Contrary to the above, since October 2020, the licensee failed to precondition equipment qualified by test by natural or artificial (accelerated) aging to its end-of-installed life condition and consider all significant types of degradation which can have an effect on the functional capability of the equipment. Specifically, the licensee failed to consider the known potential failure modes (aging) of the CHRM door gasket to determine its qualified life which can have an effect on the functional capability of the CHRM.

The licensee entered this issue into its CAP as CR 50096896 and CAP-IR-2021-6870 and is working on corrective actions to address the issue. This issue does not represent an immediate safety concern because there was no fuel in the reactor and these components were not required to perform their safety function at the time of the inspection.

Because this violation was not repetitive or willful, and was entered into the licensees corrective action program, this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy (NCV 05200025/2021005-04, Failure to Demonstrate Qualification of Containment High Radiation Monitor Door Gasket).

NCV 05200025/2021005-05, Failure to Qualify EPAs in Accordance with IEEE 317 Introduction NRC identified a Green ITAAC finding and associated violation of 10 CFR 50, Appendix B, Criterion III, for the failure to seismically or environmentally qualify the EPAs in accordance with IEEE 317, Section 6. Specifically, the licensee failed to test or analyze the as installed assembly including cable configurations, terminations, termination box, raceways, and their respective interactions.

Description On July 9, 2021, the inspectors reviewed EQDP SV3-EY01-VBR-004, Low Voltage Power, Control, and I&C Electrical Penetration Assemblies, Rev 4, to determine the methods of the seismic and environmental qualification used for the low voltage EPAs. The inspectors determined that the qualification was performed using testing combined with analysis, and the tests were conducted on a partial prototype EPA without the interfaces of the terminal boxes or raceways. In addition, the inspectors determined the tested external cable configuration was different than that of the as-installed condition. The review of the tests determined that the prototype EPA was flexible, and these interfaces could make the complete EPA more flexible with more degrees of freedom. The review of the analyses in the EQDP Section 4, Qualification by Analysis, determined the analysis used to support the testing also did not model the EPA with these interfaces either or provide justification as to why there would be no impacts. The analysis only considered a linear length of cable and connector, but did not consider the termination box, raceway, or as-installed configuration of the external cables.

The EQDP indicated that the EPAs were qualified in accordance with IEEE 317-1987.

By failing to include the interfaces in the qualification, the licensee did not meet the requirements in IEEE 317, Sections 6.1.3 and 6.2.10. These sections describe that the test configuration of the EPA be representative of the design being qualified, and that the tested configuration include consideration of terminal boxes, external cables, and raceways.

Additional analysis was performed to qualify the EPAs, which was conducted in accordance with IEEE 323 and 344-1987. The analysis failed to provide justification as why the configuration with terminal boxes, external cables, and raceways had no impact. The analysis did not represent the dynamic properties nor evaluated the nonlinear behavior that could potentially exist between the components. This was necessary to determine the responses of the complete EPA and the potential interactions of the equipment and interfaces from those responses.

In addition, the EQDP, Subsection 4.6, Maximum Unsupported Cable Length Analysis,

[APP-EY01-V7R-001, Archival of MIRION IPS-2545, EPA Analysis, Rev. 0] extended the maximum unsupported cable length for partial prototype EPAs. Based on their review, the inspectors found that the Mirion prototype model used to extend the cable length for the AP1000 EPAs did not consider or model the interfaces between the termination box, raceways, and external cables. Based on the flexibility identified during the Mirion prototype testing, the components should have been included in the tested configuration or justification provided as to why there would be no additional impacts. In addition, this analysis inappropriately analyzed test data from a similar EPA to extrapolate benchmark acceptance criteria for cable length using the static coefficient method. This was an unsuitable approach to reasonably estimate the nature and magnitude of deformations that may have occurred from the tested EPA, cables, and terminations, while implying that the extrapolated results were conservative.

The inspectors determined that the combination of testing and analysis in the EQDP brought into question the qualification of the EPA with the included interfaces. Without test, mathematical model, or an analysis that provides logical proof that the interfaces

would not impact the design or function of the EPA, the inspectors were unable to verify the seismic or environmental qualification of the EPAs. The licensee entered this issue into its CAP as CR 50104476 and performed additional analysis to show that the finding did not impact a design function of the EPAs.

Analysis The failure to seismically or environmentally qualify the EPAs in accordance with IEEE 317, Section 6.2.10 was contrary to 10 CFR 50, Appendix B, Criterion III and a performance deficiency. Per the guidance in IMC 0613, Appendix E, Examples of Minor Construction Issues, the performance deficiency was determined to be more than minor, and thus a finding, because it represented an adverse condition that rendered the quality of an SSC unacceptable or indeterminate and required substantive corrective action. Specifically, the failure to consider all significant types of degradation that could have an adverse impact on EPA functional capability for environmental and seismic qualification could allow unacceptable degradation during a design basis accident.

The inspectors determined this performance deficiency was material to the acceptance criteria of an ITAAC. The acceptance criterion of ITAAC 2.2.01.05.i requires, in part, that the as-built equipment including anchorage is seismically bounded by the tested or analyzed conditions, and a report exists and concludes that the as-built Class 1E equipment and the associated wiring, cables, and terminations identified in Table 2.2.1-1 as being qualified for a harsh environment are bounded by type tests, analyses, or a combination of type tests and analyses. The inspectors determined that the as built equipment, identified in Table 2.2.1-1, including anchorage was not seismically bounded by the tested or analyzed conditions, and the Class 1E equipment was not bounded by type tests, analyses, or a combination of type tests and analyses.

The inspectors determined the finding was associated with the Construction/Installation cornerstone of the Construction Reactor Safety strategic performance area. The inspectors assessed the finding using IMC 2519, Appendix A, AP1000 Significance Determination Process, dated October 26, 2020, and determined that the finding was not associated with a security program determined, the finding was associated with the failure to properly implement an IMC 2504 operational/construction program; and it was not associated with a repetitive, NRC-identified omission of a program critical attribute.

The inspectors determined the finding was a performance deficiency with a very low safety significance because the licensee was able to show through additional analysis that the finding did not impact a design function of the EPAs.

The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of Change Management in the area of Human Performance, in accordance with IMC 0613, Appendix F, Construction Cross-Cutting Areas and Aspects, dated November 4, 2020. Specifically, the licensee leadership was not using a systematic process for evaluating and implementing changes to as-installed SSCs compared to configurations that were environmentally qualified so that nuclear safety remained the overriding priority [H.3].

Enforcement

10 CFR 50 Appendix B, Criterion III, Design Control, required, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Since October 2020, the licensee qualification program failed to provide design control measures for verifying or checking the adequacy of EPA design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, the licensee failed to provide design control measures through testing or analysis for verifying or bounding the impacts of the termination box, raceways, terminations, their interfaces, and their interactions on the EPA assembly as whole for seismic and environmental qualification.

The licensee entered this issue into its corrective action program as CR 50104476 and is working on corrective actions to address the issue. This issue does not represent an immediate safety concern because there was no fuel in the reactor and these components were not required to perform their safety function at the time of the inspection.

Because this violation was not repetitive or willful, and was entered into the licensees CAP, this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy (NCV 05200025/2021005-05, Failure to Qualify EPAs in Accordance with IEEE 317).

3P02 Fire Protection Program

• 64705-02.01 - Implemented Operational Feature of the Fire Protection Program

• 64705-02.02 - Adequacy and operational readiness

a. Inspection Scope The inspectors reviewed aspects of the licensees fire protection program to determine if reasonable assurance existed at the time of the inspection to verify that the reviewed aspects of the program will meet the requirements of 10 CFR 50.48, Fire Protection, once fully implemented after the 10 CFR 52.103(g) finding. The inspectors reviewed the sites fire protection program document, fire hazards analysis, and fleet and site-specific procedures to determine if the requirements of Branch Technical Position CMEB 9.5-1 were incorporated into the fire protection program.

Specifically, the inspectors reviewed various aspects of the fire protection program to determine if:

• the program provided for returning the nonfunctional equipment to service in a reasonable period of time,

• the licensee performed a post-fire SSD analysis, and that the analysis demonstrated that the SSCs important to safety can accomplish their respective post-fire SSD functions,

• the sites fire protection program required all changes to the program to be reviewed and documented to determine that the change does not adversely

affect the ability to achieve and maintain safe shutdown in the event of a fire, and

• the licensees fire protection program and change process were adequate to ensure that all necessary program documents, calculations, etc. are updated as a result to any change to the program.

b. Findings No findings were identified.

3P03 Inservice Testing

• 73758-App A - Appendix A, Review of Functional Design, Qualification, and Preservice Testing/Inservice Testing (PST/IST) Programs for Pumps, Valves, and Dynamic Restraints

• 73758-App B - Appendix B, Implementation of Functional Design and Qualification Program for Pumps, Valves, and Dynamic Restraints

a. Inspection Scope The inspectors reviewed procedure NMP-ES-013-007, IST Program Data Trending and Evaluation, and verified that it was in compliance with the ASME OM Code and supports the implementation of the Vogtle 3&4 IST Program. Specifically, NMP-ES-013-007 is a Southern Nuclear Fleet procedure that was updated to add applicability to Vogtle Units 3&4.

b. Findings No findings were identified.

3P04 Preservice Inspection

• 73754-02.02 - Personnel Qualification & Certification

• 73754-02.03 - Non-destructive Examination (NDE) Review

a. Inspection Scope The inspectors observed volumetric ultrasonic NDE on the direct vessel injection A/B line to verify if the NDE was conducted in accordance with approved procedures. The inspectors also reviewed the equipment and personnel qualifications.

b. Findings No findings were identified.

3P05 Preservice Testing

• 73758-App A - Appendix A. Review of Functional Design, Qualification, and PST/IST Programs for Pumps, Valves, and Dynamic Restraints

• 73758-App B - Appendix B. Implementation of Functional Design and Qualification Program for Pumps, Valves, and Dynamic Restraints

a. Inspection Scope The inspectors performed the following activities related to the development and implementation of the Functional Design and Qualification Program for pumps, valves, and dynamic restraints that will perform safety-related functions at Vogtle Units 3 & 4:

• The inspectors reviewed motor-operated valve (MOV) design specifications, application report, functional qualification report, and other documents, related to the qualification of a sample of motor-operated butterfly valves at Vogtle Units 3 & 4. The inspectors discussed the MOV qualification activities described in these documents with licensee staff and contractors. On this basis, the inspectors determined that the licensee satisfied the design specification provisions for implementing ASME Standard QME-1-2007, which the NRC accepted in Regulatory Guide 1.100 (Revision 3), for the MOVs listed below:

o NPS 10 Class 150 TRICENTRIC Butterfly Valves with Limitorque Motor Actuators

• The inspectors reviewed design specifications, application reports, functional qualification reports, equipment qualification reports and data packages, and other documents, related to the qualification of a sample of pyrotechnic-actuated (squib) valves at Vogtle Units 3 & 4. The inspectors discussed the qualification activities described in these documents with licensee staff and contractors. For example, the inspectors verified that the qualification analysis addressed the capability of the piping system to withstand the combined loads (including end loading) during actuation of the squib valves. Further, the inspectors confirmed that the post-installation performance of the squib valves is addressed by lot testing of active parts of the squib valves, such as actuator cartridges, tension bolts, and shear caps. On this basis, the inspectors determined that the licensee satisfied the design specification provisions for implementing ASME Standard QME-1-2007, which the NRC accepted in Regulatory Guide 1.100 (Rev. 3), for the squib valves listed below:

o 8-inch Class 150 and Class 2500 Squib Valves o 14-inch Class 2500 Squib Valves

b. Findings No findings were identified.

3P06 Reactor Vessel Material Surveillance

• 50054-02.02 - Specific Requirements

a. Inspection Scope The inspectors performed a walkdown of the capsule brackets to verify if they were installed in accordance with the design. The inspectors observed the locations of the capsule brackets to verify if they were installed in accordance with the design drawings. The inspectors also visually reviewed the vessel material samples to verify if the samples were the material specified in the design specification. The inspectors observed the installation of the vessel material samples to verify if they were installed in accordance with the installation procedure.

b. Findings No findings were identified.

4. OTHER INSPECTION RESULTS 4OA5 Other Activities

.01

a. Inspection Scope The inspectors reviewed the corrective actions associated NCV 0520025/2021002-02, Failure to Use Worst Case Load Profile for IDS Battery Service Test, (CR 50080247) to verify if the violation was corrected. Specifically, the inspectors reviewed 3-IDS-ITPP-501, "Class 1E DC and UPS Preoperational Testing," to verify if the worst-case load profile was used for the battery service tests.

b. Findings No findings were identified.

.02

a. Inspection Scope The inspectors reviewed corrective actions developed and implemented by the licensee to verify if the violation identified as05200025/2020009-01, Failure to Complete Containment Prior to Unit 3 ILRT, was corrected.

b. Findings No findings were identified.

4OA6 Meetings, Including Exit

.1 Exit Meeting.

On October 18, 2021, the inspectors presented the inspection results to Mr. G.

Chick, Vogtle 3&4 Executive Vice President, and other licensee and contractor staff members. Proprietary information was reviewed during the inspection period but was not included in the inspection report.

SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensees and Contractor Personnel A. Nix, NI Manager K. Roberts, ITAAC Manager M. Hickox, Test Support Manager C. Alexander, Milestone Manager S. Boyle, Milestone Manager D. Pagan-Diaz, ITP Turnover. Manager J. Olsen, NI Supervisor N. Kellenberger, SNC Licensing Supervisor C. Castell, SNC Licensing Engineer N. Patel, SNC Licensing Engineer J. Cole, SNC Licensing Engineer J. Weathersby, SNC Licensing Engineer C. Main, ITAAC Project Manager D. Wade, ITAAC Project Manager B. Macioce, Principle Engineer Digital Testing R. McKay, ITP Test Engineer S. Turner, ITP Test Engineer G. Weaver, ITP Test Engineer R. Nicoletto, ITP Test Engineer W. Pipkins, ITP Test Engineer D. Melton, ITP Test Engineer R. Espara, ITP Test Engineer J. Clark, ITP Test Engineer K. Morgan, ITP Test Engineer LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Item Number Type Status Description 05200025/2021005-01 NCV Open/Closed Failure to Follow Procedure 3-IDSC-SOP-002 05200025/2021005-02 NCV Open/Closed Failure to Obtain a License Amendment Prior to Altering Dynamic Restraint Qualification Requirements05200025/2021005-03 NCV Open Failure to Demonstrate Qualification of Valve Nonmetallic Parts05200024/2021005-04 NCV Open Failure to Demonstrate Qualification of Containment High Radiation Monitor Door Gasket 05200025/2021005-05 NCV Open Failure to Qualify EPAs in Accordance with IEEE 317 05200025/2020009-01 NCV Closed Failure to Complete Containment Prior to Unit 3 ILRT 05200025/2021002-02 NCV Closed Failure to Use Worst Case Load Profile for IDS Battery Service Test LIST OF DOCUMENTS REVIEWED Section 1 Section 1A01 APP-VAS-MD-659, Auxiliary Building Area 5 Elevation 153'0" VAS Duct ISO View, Rev. 9 APP-MD03-V2-850003, DFDAF-330 Damper Schedule, Rev. 15 APP-MD03-V2-850006, Fire Damper Grating Assembly, Rev. 1 APP-AB01-AB-010, Blockouts and Barriers (Penetrations, Seals and Fire Stops) Details Sheet 10, Rev. 4 APP-AB01-AB-012, Blockouts and Barriers (Penetrations, Seals and Fire Stops) Details Sheet 12, Rev. 0 Engineering & Design Coordination Report No. APP-FSAR-GEF-061, Addition of AS04 Specialty Device Detail to HVAC, Rev. 0 Engineering & Design Coordination Report No. APP-SES-GEF-031, Updates to Barrier Matrix, Rev. 0 APP-AS21-A1-001, AP1000 Security Barrier Design Requirements, Rev. 1 APP-GW-MD-103, HVAC Details Sheet 1, Rev. 1 Section 1P01 NMP-AP-003, Procedure and Work Instruction Use and Adherence, Version 6.1 3-IDSC-SOP-002, Class 1E AC System - Division C, Version 0

3. OPERATIONAL READINESS Section 3T01 ND-21-0627, ITAAC Closure Notification on Completion of Item 2.1.02.08b [Index Number 30],

08/2021 SV3-RCS-ITR-800040, Unit 3 Recorded Results of RCS Flow Coastdown Flow Measurement Test: ITAAC 2.1.02.08b, NRC Index Number: 30, 08/2021 SV3-RCS-T2C-5063, Vogtle Unit 3 RCS Flow Coastdown Preoperational Testing Results Validation, Rev. 0

SV3-RCS-T2R-5063, Vogtle Unit 3 RCS Flow Coastdown Preoperational Testing Summary Report, Rev. 0 Work Order 1071744 2.1.02.08b-U3-CP, ITAAC Completion Package, Rev. 0 Section 3T02 3-RCS-ITPP-506, Reactor Coolant Pump and Reactor Coolant Flow Precore Hot Functional, Ver. 3.0 SV3-RCS-T2C-5061, Component, Hot Leg, Elbow, and Cold Leg Bend Differential Pressure Flow Calculation from Vogtle Unit 3 Hot Functional Testing, Rev. 0 SV3-RCS-T2R-5061, Vogtle Unit 3 Hot Functional Testing - Reactor Coolant System Flow Summary Report, Rev. 0 ND-21-0632, ITAAC Closure Notification on Completion of 2.1.02.09a [Index Number 41],

08/2021 SVP-SV0-006408, Completion Package for ITAAC 2.1.02.09a , 07/2021 Section 3T03 ND-20-0749, ITAAC Closure Notification on Completion of ITAAC 2.1.02.11a [Index Number 47], 08/2021 ITAAC Technical Reports:

SV3-RCS-ITR-800047, Unit 3 Recorded Results of Remotely Operated RCS Valves Controlled by PMS and the Main Control Room: ITAAC 2.1.02.11a.ii Item 11.a and 11.b, NRC Index Number: 47, Rev. 0 SV3-RCS-ITR-801047, Unit 3 Recorded Results of Remotely Operated RCS Valves Response to Loss of Motive Power: ITAAC 2.1.02.11a.ii Item 12b, NRC Index Number: 47, Rev. 0 Work Package SV3-RCS-T0W-SNC921644, Perform ITAAC 2.1.02.11a.ii Item 10 2.1.02.11a.ii-U3-CP, ITAAC Completion Package, Rev. 0 B-GEN-ITPCI-001, PMS Cabinets, Ver. 1.2 Section 3T04 3-PXS-ITPP-505, ADS Stages 1-3 Dynamic Test, Rev. 1.1 WO 1071720 ND-21-0625 ITAAC Closure Notification on Completion of 2.1.02.12a.iii [Index Number 55],

07/2021 2.1.02.12aiii-U3-CP, ITAAC Completion Package, Rev. 0 Section 3T05 SV3-CVAP-T2R-100, Vogtle Unit 3 Pre-Hot Functional Test Inspection of Reactor Vessel Internals, Rev. 3.0 3-RXS-ITPP-501, Pre- and Post-Hot Functional Test Inspection of Reactor Vessel Internals, Rev. 3.0 WO 1061754 CR 50104315 CR 50103428 TE 60029420 ESR 50104314

Section 3T06 3-CCS-ITPP-501, Component Cooling Water System Preoperational Test Procedure, Rev. 2.0 3-CNS-ITPP-503, LLRT Containment Leak Rate Test Type C, Rev. 3.0 WO 1048157SV3-CNS-ITR-800116, Unit 3 Recorded Results of CCS Motor-Operated Valves Change Position as Indicated in Table 2.21-1: ITAAC 2.2.01.11a.iii, Rev. 0 G-GEN-ITPCM-001, Limitorque SMB/SB Motor Operated Valve Component Testing, Ver. 3.0 SV3-CNS-ITR-801116, Unit 3 Recorded Results of SFS Motor-Operated Valves Change Position as Indicated in Table 2.21-1: ITAAC 2.2.01.11a.iii, Rev. 0 3-SFS-ITPP-502, Spent Fuel Pool Cooling System Flow Path Preoperational Test Procedure, Ver. 3.1 Section 3T07 ND-21-0745, ITAAC Closure Notification on Completion of ITAAC 2.2.01 .11 b [Index Number 118], 08/2021 ITAAC Technical Reports (ITRs):

SV3-CNS-ITR-800118, Unit 3 Test Results for CAS Containment Isolation Valves Loss of Motive Power Testing: ITAAC 2.2.01.11 b, NRC Index Number: 118, Rev. 1 SV3-CNS-ITR-801118, Unit 3 Test Results for CCS Containment Isolation Valves Loss of Motive Power Testing: ITAAC 2.2.01.11 b, NRC Index Number: 118, Rev. 0 SV3-CNS-ITR-802118, Unit 3 Test Results for SFS Containment Isolation Valves Loss of Motive Power Testing: ITAAC 2.2.01.11 b, NRC Index Number: 118, Rev. 0 SV3-CNS-ITR-803118, Unit 3 Test Results for VFS Containment Isolation Valves Loss of Motive Power Testing: ITAAC 2.2.01.11 b, NRC Index Number: 118, Rev. 1 SV3-CNS-ITR-804118, Unit 3 Test Results for VWS Containment Isolation Valves Loss of Motive Power Testing: ITAAC 2.2.01.11 b, NRC Index Number: 118, Rev. 1 SV3-CNS-ITR-805118, Unit 3 Test Results for WLS Containment Isolation Valves Loss of Motive Power Testing: ITAAC 2.2.01.11 b, NRC Index Number: 118, Rev. 1 2.2.01 .11 b-U3-CP, ITAAC Completion Package, Rev. 0 Section 3T08 3-PXS-ITPP-504, Passive Core Cooling System Hot Functional Test, Rev. 2.1 3-GEN-ITPP-517, Precore Hot Functional Test Procedure, Rev. 5.0 WO 1071719 2.2.03.08b.01-U3-CP, ITAAC Completion Package, Rev. 0 ND-21-0868, ITAAC Closure Notification on Completion of ITAAC 2.2.03.08b.01 [NRC Index Number 175], 09/2021 SV3-PXS-T2C-011, Vogtle Unit 3 PXS Hot Functional Test Results Validation for PRHR Performance, Rev. 0 SV3-PXS-ITR-800175, Unit 3 Recorded Results of PRHR Heat Exchanger Heat Transfer Rate Test: ITAAC 2.2.03.08b.01, Rev. 0 Section 3T09 B-GEN-ITPCE-008, "Class 1E and Non-Class 1E Battery testing,"(Divisions A and B 125 VDC 24hr Battery 1B only), Ver. 7 APP-DB01-Z0-001, "Design Specification for Class 1E 250 VDC Batteries and Racks," Rev. 10 3-IDS-ITPP-501, "Class 1E DC and UPS Preoperational Testing," Ver. 5.1 CR 50080247

Section 3P01 APP-GW-G0X-003, AP1000 Commodity Locator Codes, Rev. 11 APP-GW-G1X-001, Governing AP1000 Design Codes and Standards, Rev. 10 APP-GW-VP-010, Equipment Qualification Methodology and Documentation Requirements for AP1000 Safety-Related Valves and Valve Appurtenances, Rev. 3 CR 50098695 LDCR-2016-107, Snubber Qualification Changes, Rev. 2 LDCR-2016-107, Snubber Qualification Changes, Rev. 1 APP-GW-VPD-002, AP1000 Safety Related Component List, Rev. 11 APP-GW-G1-002, AP1000 Plant Equipment Qualification Methodology, Rev. 3 APP-GW-N2C-009, Beta Reduction Factors and Gamma Equivalency for Various Shielding Materials, Rev. 0 APP-PV70-T5-001, Qualification Plan for Safety-Related Squib Valve Actuators, Electrical Connector Assemblies, and Bracket Assemblies for Westinghouse Electric Company for use in Westinghouse AP1000 Nuclear Power Plants, Rev. 5 LTR-EQ-18-23, Activation Energy for TE Connectivity Splice Materials, 05/22/2018 ND-AD-002, Nuclear Development Corrective Action Program, Rev. 29.1 ND-EN-VNP-006, Equipment Qualification Data Package (EQDP) Review and Acceptance, Rev. 4.0 ND-LI-VNP-002, Applicability Determination and 50.59 / Departure Screening for VEGP 3&4, Rev. 18 ND-LI-VNP-002-F06, 50.59 / Departure Screening for Vogtle 3&4, Rev. 6 ND-LI-VNP-003, 50.59 / Departure Evaluations for Vogtle 3&4, Rev. 8.0 ND-LI-VNP-007, Licensing Document Change Requests for VEGP 3&4, Rev. 10 NMP-AD-008, Applicability Determinations, Rev. 21.1 SV0-EY20-VPC-001, Vogtle Units 3 and 4 EY20 Cale Splice Thermal Aging and Post-DBA Aging Calculation, Rev. 0 APP-PV64-Z0-001, Design Specification Main Steam Isolation, ASME BPVC,Section III, Class 2, Rev. 7 APP-PV70-Z0-001, Squib (Pyrotechnic Actuated) Valves, ASME Boiler and Pressure Vessel Code Section III, Class 1, Rev. 7 App-SS30-Z0-002, Design and Fabrication Requirements for Hydraulic Shock Absorbers (Snubbers), Rev. 4 APP-EY20-VBR-002, Equipment Qualification Data Package for Class 1E Low Voltage Cable Splices for Use in the AP1000 Plant, Rev. 0 APP-PV64-VBR-012, Equipment Qualification Data Package for Main Steam Isolation Valves for Use in the AP1000 Plant, Rev. 0 APP-PV70-VBR-005, Equipment Qualification Data Package for 14 Squib Valves for Use in the AP1000 Plant, Rev. 1 SV3-1050-P2-0014, Nuclear Island Containment and Shield BLDG General Arrangement Operating Deck EL 135'-3, Rev. 5 SV3-PMS-VPR-009, Mild Environment Abnormal Temperature and Relative Humidity Extremes Report for NIS and RMS Cable Connection Assemblies for Use in the AP1000 Plant, Rev. 0 SV3-EJ01-VBR-002, Equipment Qualification Data Package for EJ01 Class 1E Junction Boxes for Use in the AP1000 Plant, Rev. 0 SV3-EY20-VBR-002, Low Voltage Cable Splices, Rev. 0 SV3-1050-P2-0014 Nuclear Island Containment and Shield BLDG General Arrangement Operating Deck EL 135'-3, Rev. 5

SV3-PMS-VPR-009 Mild Environment Abnormal Temperature and Relative Humidity Extremes Report for NIS and RMS Cable Connection Assemblies for Use in the AP1000 Plant, Rev. 0 SV3-RMS-V0-001 AP1000 Radiation Monitoring System Vendor Outline Drawing Package, Rev. 1 SV3-RMS-VBR-002 Equipment Qualification Data Package for the Radiation Monitoring System (RMS) Containment High Range Monitor (CHRM) and Field Cable Connections at the EPA Feedthroughs, Rev. 0 SV3-RMS-VBR-007 Containment High Range Monitor Design Basis Accident Test Report, Rev.

0 Section 3P02 Drawings APP-1000-AF-901, Fire Area Drawing Nuclear Island Section AA, Rev. 5 APP-AB01-AB-001, Blockouts and Barriers (Penetrations, Seals and Fire Stops) Details-Sheet 1, Rev. 9 SV3-1210-JFH 800500, Fiber Optic Linear Heat Detection Layout and Wiring, Rev. 1 57811-BMT-11306-AW-138, Containment Building Room 11306, Elevation 107-2, MSW WO3/WO5 L500 Board Layout, Rev.1 SV3-1200-CRK-CV0893, Unit 3 Auxiliary Building Walls El. 66 6 Concrete Placement Layout, Rev. 11 SV3-1230-CCK-CV4261, Unit 3 Auxiliary Building Wall Placements from El. 100-0 to El. 117-6, Rev. 1 SV3-1220-CRK-CV2157, Unit 3 Auxiliary Building Walls El. 82-6 to 100-0 Concrete Placement Layout, Rev. 3 APP-1010-AF-001, Fire Area Drawing Nuclear Island Plan at El. 66-6, Rev. 5 APP-1020-AF-001, Fire Area Drawing Nuclear Island Plan at El. 82-6, Rev. 6 APP-1020-AF-002, Fire Area Drawing Nuclear Island Plan at El. 96-6, Rev. 6 APP-1030-AF-001, Fire Area Drawing Nuclear Island Plan El. 100-0 and 107-2, Rev. 6 APP-1040-AF-001, Fire Area Drawing Nuclear Island Plan at El. 117-6, Rev. 5 APP-IDSA-E3-DK-101, One line Diagram Class 1E 250V DC MCC IDSA-DK-1 Auxiliary Bldg, Sheet 1, Rev. 5 APP-IDSA-E3-DK-102, One line Diagram Class 1E 250V DC MCC IDSA-DK-1 Auxiliary Bldg, Sheet 2, Rev. 4 APP-IDSA-E3-DK-103, One line Diagram Class 1E 250V DC MCC IDSA-DK-1 Auxiliary Bldg, Sheet 3, Rev. 1 APP-IDS-E3-001, Class 1E DC System Station One Line Diagram Divisions A & C, Sheet 1, Rev. 3 APP-IDS-E3-002, Class 1E DC System Station One Line Diagram Divisions B, D & Spare, Sheet 1, Rev. 3 Licensing & Design Basis Docs APP-GW-N4R-003, Fire Protection Analysis Report, Rev. 2 NMP-ES-035, Fire Protection Program, Ver. 7.0 NMP-ES-035, Fire Protection Program Implementation, Ver. 14.0 NMP-ES-035-003, Fleet Hot Work Instructions, Ver. 10.0 NMP-ES-035-014, Fleet Transient Combustible Controls, Ver. 5.0 NMP-ES-035-019, Pre-Fire Plans, Ver. 2.0 APP-GW-N4R-003, Fire Protection Analysis Report, Rev. 2 APP-FPS-M3-001, AP1000 Fire Protection System (FPS) System Specification Document, Rev. 0

COL VEGP Combined License Vogtle Electric Generating Plant Unit 3, Through Amendment 186 VEGP 3 & 4, UFSAR, Rev. 10 Vogtle Units 3 & 4 Technical Requirements Manual (TRM), Rev. 18 SVO-SES-EF-900009, Distributed Antenna System Design, Rev. 2 ND-18-1408, Request for License Amendment Regarding Routing of Class 1E Divisional Cables Supporting Passive Containment Cooling (LAR-18-028), 11/16/2018 ND-19-0041, Supplement to Request for License Amendment Regarding Routing of Class 1E Divisional Cables Supporting Passive Containment Cooling (LAR-18-028S1), 1/24/2019 Procedures 3-AOP-115, Loss of Normal Residual Heat Removal, Rev. H.7 3-EOP-SDP-2, Response to Loss of RNS During Shutdown, Rev. F.5 3-AOP-601, Evacuation of Control Room, Rev. J.9 3-AOP-902, Fire Response Emergency, Rev. H.7 B-GEN-OPS-005, Fire Response Procedure, Rev. 5 B-GEN-OPS-004, Fire Brigade Equipment Quarterly Inspection, Ver. 2.0 B-PFP-ENG-001, Control of Pre-Plans, Ver. 3.0 B-PFP-ENG-001-F3110, Aux. Bldg. non-RCA, El 66-6 B-PFP-ENG-001-F3114, Aux. Bldg. non-RCA, El 117-6 B-PFP-ENG-001-F3128, Aux. Bldg. & PCS Valve Room B-GEN-ENG-008, Fire Equipment Functionality and Fire Protection Impairments (FPI)

Requirements, Version 1.0 B-GEN-ENG-008-GL02, Fire Equipment Functionality and Fire Protection Impairments (FPI)

Basis, Version 1.0 ND-LI-NP-007, Licensing Document Change Requests for VEGP Units 3 & 4, Version 9.0 NMP-ES-035-005, Fire Protection Alternative Compensatory Measures, Version 6.0 NMP-ES-035-006, Fire Protection Program Impact Screen and Detailed Reviews, Version 11.0 NMP-ES-035-006-F05, Fire Protection Program Impact Screen, Version 5.0 NMP-ES-035-006-F06, Detailed Fire Protection Program Change Evaluation, Version 4.0 NMP-ES-095-001, Documentation Change Processes, Version 1.0 Work Orders SNC921209, FSP Individual Fire Detector Test Package SV3-1230-CCW-CV2443, Aux Building Battery Rack Walls (32, 33, 34, 36, & 37) up to El. 100-0, closed 10/29/2020 SV3-1230-CCW-800001, 1084 to 115-6, CL Wall, Concrete Placement & Post-Placement (Wall 89), closed 2/7/2019 Miscellaneous 3-AOP-902-B, Background Information for 3-AOP-902, Fire Response Emergency, Rev. G.6 APP-IDS-E8-001, Class 1E DC and UPS System Specification Document, Rev. 5 B-017-CCR 100318946, Fire Brigade Training Roster, May 12, 2021 B-017-CCR 100319184, Fire Brigade Training Roster, May 19, 2021 Vogtle 3 & 4 Fire Protection Regulation Structure (Applicable NFPA Codes of Record), Revised 6/25/2021 FPRA Calculation List, Revised 6/25/2021 Maximo FPS PM List, Revised 6/25/2021 Fire Pre-Plan Index and Status List, Revised 6/25/2021 SV3-1200-ITR-800797, Fire Barrier Inspection Report, Unit 3: ITAAC 3.3.00.07c.ii.a, 3.3.00.07c.ii.b, Rev. 0

VS3-ITAAC-ST-2.3.04.10, FPS Individual Fire Detectors - ITAAC: SV3-2.3.04.10, Rev. 0 WCAP-15871, AP1000 Assessment Against NFPA 804, Rev. 1 Corrective Action Documents Reviewed CR 50092969, Equivalent Composite Firewall material not part of a listed assembly, 5/18/2021 CR 50004139, Cable Route Design Discrepancy with COL Appendix C, 8/31/2018 CAR 50007691, Cable Route Design Discrepancy with COL Appendix C, 8/31/2018 TE 60003574, Cable Route Design Discrepancy with COL Appendix C, 8/31/2018 Condition Reports Generated During Inspection CR 50103154, WCAP-15871 incorrect statement on listing of fire walls and penetrations, 8/10/2021 Section 3P03 Procedure NMP-ES-013-007, IST Program Data Trending and Evaluation, Revision 3.0 Section 3P05 Design Specifications APP-PV11-Z0-001, Design Specification for Butterfly Valves, ASME Boiler and Pressure Vessel Code Section III Class 2 and 3, Rev. 13 APP-PV70-Z0-001, Squib (Pyrotechnic Actuated) Valves, ASME Boiler and Pressure Vessel Code Section III, Class 1, Rev. 7 Datasheets APP-PV11-Z0D-122, PV Datasheet 122, Rev. 7 Calculations APP-GW-PVR-002, Piping and Valve Interface Requirements Document, Rev. 3 APP-GW-VP-010, Equipment Qualification Methodology and Documentation Requirements for AP1000 Safety-Related Valves and Valve Appurtenances, Rev. 3 APP-PV11-VPR-003, Functional Qualification Report for a NPS 8 TRICENTRIC Butterfly Valve, Rev. 0 APP-PV70-T5-001, Qualification Plan for Safety-Related Squib Valve Actuators, Electrical Connector Assemblies, and Bracket Assemblies for Westinghouse Electric Company for use in Westinghouse AP1000 Nuclear Power Plants, Rev. 5 APP-PV70-VBR-002, Equipment Qualification Summary Report for 8 Squib Valves for Use in the AP1000 Plant, Rev. 1 APP-PV70-VBR-003, Equipment Qualification Data Package for 8 Squib Valves for Use in the AP1000 Plant, Rev. 1 APP-PV70-VBR-004, Equipment Qualification Summary Report for 14 Squib Valves for Use in the AP1000 Plant, Rev. 1 APP-PV70-VBR-005, Equipment Qualification Data Package for 14 Squib Valves for Use in the AP1000 Plant, Rev. 1 APP-PV70-VBR-005, Equipment Qualification Data Package for 14 Squib Valves for Use in the AP1000 Plant, Rev. 1 APP-PV70-VPR-013, ASME QME-1-Functional Qualification Report for 14 Squib Valves Installed in AP1000 Plants, Rev. 0

APP-PV70-VPR-014, ASME QME-1 Functional Qualification Report for 8 Squib Valves Installed in AP1000 Plants, Rev. 0 APP-PV70-VPR-016, QME-1 Testing of a 14-2500# Squib Valve for Westinghouse Electric Company, LLC Job Number T57622 (PR028955), Rev. 0 APP-PV70-VPR-021, Test Report of QME Testing Performed on 8 inch HP Squib Valve, Re. 0 APP-PV70-VQQ-002, 14 ADS Squib Valve Assembly Data Package, 08/2018 APP-PV70-VQQ-053, Lot Acceptance Testing, Rev. 0 APP-PXS-PLR-010, AP1000 Direct Vessel Injection Line A (APP-PXS-PLR-010) Piping Stress Analysis Report, Rev. 6 APP-PXS-PLR-020, AP1000 Direct Vessel Injection Line B (APP-PXS-PLR-020) Piping Stress Analysis Report, Rev. 6 APP-PXS-PLR-030, Piping Analysis Report for Loop 1 - Automatic Depressurization System 4th Stage West Compartment and Passive Residual Heat Removal Supply, Rev. 4 APP-RCS-PLR-030, Piping Analysis Report for Reactor Coolant System (RCS) Loop 2 -

Automatic Depressurization System (ADS) 4th Stage East Compartment, Containment Building Room 11302, Rev. 4 SV0-PV11-VPR-122 (Revision 0, 7-24-2020), Vogtle 3&4 Application Report for Tricentric NPS 10 Motor-Operated Butterfly Valves Procedure NMP-ES-013-007, IST Program Data Trending and Evaluation, Revision 3.0 Section 3P06 Drawing APP-MI01-V2-133, Reactor internals core barrel assembly, Rev. 3 Specifications APP-MI01-Z0-001, AP1000 Reactor internals functional specification, Rev. 5 APP-MI01-Z0-101, AP1000 Reactor vessel internals design specification, Rev. 10 APP-MI01-Z0-600, AP1000 Reactor vessel internals fabrication requirements, Rev. 3 QA document QR-14-3343, Quality release and certificate of conformance, Rev. 0

4. OTHER INSPECTION RESULTS Section 4OA5

.01 B-GEN-ITPCE-008, "Class 1E and Non-Class 1E Battery Capacity Testing," Rev. 7 APP-DB01-Z0-001, "Design Specification for Class 1E 250 VDC Batteries and Racks," Rev. 10 3-IDS-ITPP-501, "Class 1E DC and UPS Preoperational Testing," Ver. 5.1 CR 50080247

LIST OF ACRONYMS 10 CFR Title 10 of the Code of Federal Regulations ASME American Society of Mechanical Engineers CAP Corrective Action Program CHRM containment high range monitor COL Combined License CR condition report EPA electrical penetration assembly EQ environmental qualification EQDP equipment qualification data package EUT equipment under test IDS Class 1E DC and UPS system IEEE Institute of Electrical and Electronics Engineers IMC inspection manual chapter IP inspection procedure IR inspection report IST inservice testing ITAAC inspection, tests, analysis, and acceptance criteria LOCA loss of coolant accident MSI main steam isolation MOV motor operated valve NCV noncited violation NDE nondestructive examination NRC Nuclear Regulatory Commission PST preservice testing QA quality assurance Rev. revision SNC Southern Nuclear Company SSC systems, structures, and components UFSAR Updated Final Safety Analysis Report Ver version

ITAAC INSPECTED No. ITAAC No. Design Commitment Inspections, Tests, Acceptance Criteria Analysis 30 2.1.02.08b 8.b) The RCPs have a A test will be The pump flow rotating inertia to performed to coastdown will provide RCS flow determine the pump provide RCS flows coastdown on loss of flow coastdown curve. greater than or power to the pumps. equal to the flow shown in Figure 2.1.2-2, Flow Transient for Four Cold Legs in Operation, Four Pumps Coasting Down.

41 2.1.02.09a 9.a) The RCS provides Testing and analysis The calculated post-circulation of coolant to to measure RCS flow fuel load RCS flow remove heat from the with four reactor rate is > 301,670 core. coolant pumps gpm.

operating at noload RCS pressure and temperature conditions will be performed. Analyses will be performed to convert the measured pre-fuel load flow to postfuel load flow with 10percent steam generator tube plugging.

No. ITAAC No. Design Commitment Inspections, Tests, Acceptance Criteria Analysis 47 2.1.02.11a.ii 10. Safety-related Inspection will be Safety-related displays identified in performed for displays identified in Table 2.1.2-1 can be retrievability of the Table 2.1.2-1 can be retrieved in the MCR. safety-related displays retrieved in the 11.a) Controls exist in in the MCR. ii) MCR. ii) Controls the MCR to cause the Stroke testing will be in the MCR operate remotely operated performed on the to cause the valves identified in other remotely remotely operated Table 2.1.2-1 to operated valves listed valves (other than perform active in Table 2.1.2-1 using squib valves) to functions. 11.b) The controls in the MCR. perform active valves identified in ii) Testing will be functions. ii) The Table 2.1.2-1 as having performed on the other remotely PMS control perform an other remotely operated valves active safety function operated valves identified in Table after receiving a signal identified in Table 2.1.2-1 as having from the PMS. 12.b) 2.1.2-1 using real or PMS control perform After loss of motive simulated signals into the active function power, the remotely the PMS. iii) Testing identified in the table operated valves will be performed to after receiving a identified in Table demonstrate that signal from PMS. iii) 2.1.2-1 assume the remotely operated These valves open indicated loss of motive RCS valves within the following power position. RCSV001A/B, times after receipt of V002A/B, V003A/B, an actuation signal:

V011A/B, V012A/B, V001A/B < 40 sec V013A/B open within V002A/B, V003A/B the required response < 100 sec V011A/B times. Testing of the < 30 sec remotely operated V012A/B, V013A/B valves will be < 60 sec Upon loss performed under the of motive power, conditions of loss of each remotely motive power. operated valve identified in Table 2.1.2-1 assumes the indicated loss of motive power position.

55 2.1.02.12a.iii 12.a) The automatic iii) Tests of the motor- iii) Each motor-depressurization valves operated valves will operated valve identified in Table be performed under changes position as 2.1.2-1 perform an pre-operational flow, indicated in Table active safety-related differential pressure 2.1.2-1 under pre-function to change and temperature operational test position as indicated in conditions. conditions.

the table.

No. ITAAC No. Design Commitment Inspections, Tests, Acceptance Criteria Analysis 78 2.1.03.07.i 7. The reactor internals i) Not used per i) Not used per will withstand the Amendment No. 151. Amendment No.

effects of flow induced ii) A pre-test 151. ii) The as-built vibration. 10. The inspection, a flow test reactor internals reactor lower internals and a post-test have no observable assembly is equipped inspection will be damage or loose with holders for at least conducted on the as- parts. At least eight eight capsules for built reactor internals. capsules are in the storing material Inspection of the reactor lower surveillance specimens. reactor lower internals internals assembly.

assembly for the presence of capsules will be performed.

107 2.2.01.07.i 7. The CNS provides i) A containment i) The leakage rate the safety-related integrated leak rate from containment for function of containment test will be performed. the integrated leak isolation for rate test is less than containment boundary La.

integrity and provides a barrier against the release of fission products to the atmosphere.

116 2.2.01.11a.iii 11.a) The motor- iii) Tests of the motor- iii) Each motor-operated and check operated valves will operated valve valves identified in be performed under changes position as Table 2.2.1-1 perform preoperational flow, indicated in Table an active safety-related differential pressure, 2.2.1-1 under pre-function to change and temperature operational test position as indicated in conditions. conditions.

the table.

118 2.2.01.11b 11.b) After loss of Testing of the After loss of motive motive power, the remotely operated power, each remotely operated valves will be remotely operated valves identified in performed under the valve identified in Table 2.2.1-1 assume conditions of loss of Table 2.2.1-1 the indicated loss of motive power. assumes the motive power position. indicated loss of motive power position.

No. ITAAC No. Design Commitment Inspections, Tests, Acceptance Criteria Analysis 175 2.2.03.08b.01 8.b) The PXS provides 1. A heat removal 1. A report exists core decay heat performance test and and concludes that removal during design analysis of the PRHR the PRHR HX heat basis events. HX will be performed transfer rate with the to determine the heat design basis transfer from the HX. number of PRHR For the test, the HX tubes plugged reactor coolant hot leg is: 8.46 x 107 temperature will be Btu/hr with 250°F initially at 350°F with HL Temp and an the reactor coolant initial IRWST pumps running. The temperature of 80°F.

IRWST water level for The heat transfer the test will be above rate measured in the the top of the HX. The test should be test will continue until adjusted to account the hot leg for differences in the temperature is HL and IRWST 250°F. temperatures and the number of tubes plugged.

603 2.6.03.04c 4.c) Each IDS 24-hour Testing of each 24- The battery terminal battery bank supplies a hour asbuilt battery voltage is greater dc switchboard bus bank will be than or equal to 210 load for a period of 24 performed by applying V after a period of hours without a simulated or real no less than 24 recharging. 4.d) Each load, or a combination hours with an IDS 72-hour battery of simulated or real equivalent load that bank supplies a dc loads which envelope equals or exceeds switchboard bus load the battery bank the battery bank for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> design duty cycle. design duty cycle without recharging. The test will be capacity. The 4.e) The IDS spare conducted on a battery terminal battery bank supplies a battery bank that has voltage is greater dc load equal to or been fully charged than or equal to 210 greater than the most and has been V after a period of severe switchboard bus connected to a battery no less than 72 load for the required charger maintained at hours with an period without 270+/-2 V for a period equivalent load that recharging. 4.f) Each of no less than 24 equals or exceeds IDS 24-hour inverter hours prior to the test. the battery bank supplies its ac load. Testing of each 72- design duty cycle 4.g) Each IDS 72-hour hour as-built battery capacity. The inverter supplies its ac bank will be battery terminal load. 4.h) Each IDS performed by applying voltage is greater 24-hour battery charger a simulated or real than or equal to 210 provides the PMS with load, or a combination V after a period with two loss-of-ac input of simulated or real a load and duration voltage signals. 5.a) loads which envelope that equals or Each IDS 24-hour the battery bank exceeds the most battery charger design duty cycle. The severe battery bank supplies a dc test will be conducted design duty cycle switchboard bus load on a battery bank that capacity. Each 24-while maintaining the has been fully hour inverter corresponding battery charged and has been supplies a line-to-charged. 5.b) Each connected to a battery line output voltage IDS 72-hour battery charger maintained at of 208 +/- 2% V at a charger supplies a dc 270+/-2 V for a period frequency of 60 +/-

switchboard bus load of no less than 24 0.5% Hz. Each 72-while maintaining the hours prior to the test. hour inverter corresponding battery Testing of the as-built supplies a line-to-charged. 5.c) Each IDS spare battery bank will line output voltage regulating transformer be performed by of 208 +/- 2% V at a supplies an ac load applying a simulated frequency of 60 +/-

when powered from the or real load, or a 0.5% Hz. Two PMS 480 V MCC. 6. Safety- combination of input signals exist related displays simulated or real from each 24-hour identified in Table loads which envelope battery charger 2.6.3-1 can be retrieved the most severe of the indicating loss of ac in the MCR. 11. division batteries input voltage when Displays of the design duty cycle. The the loss-of-input parameters identified in test will be conducted voltage condition is

Table 2.6.3-2 can be on a battery bank that simulated. Each 24-retrieved in the MCR. has been fully hour battery charger charged and has been provides an output connected to a battery current of at least charger maintained at 150 A with an output 270+/-2 V for a period voltage in the range of no less than 24 210 to 280 V. Each hours prior to the test. 72-hour battery Testing of each 24- charger provides an hour as-built inverter output current of at will be performed by least 125 A with an applying a simulated output voltage in the or real load, or a range 210 to 280 V.

combination of Each regulating simulated or real transformer supplies loads, equivalent to a a line-to-line output resistive load greater voltage of 208 +/- 2%

than 12 kW. The V. Safety-related inverter input voltage displays identified in will be no more than Table 2.6.3-1 can be 210 Vdc during the retrieved in the test. Testing of each MCR. Displays 72-hour as-built identified in Table inverter will be 2.6.3-2 can be performed by applying retrieved in the a simulated or real MCR.

load, or a combination of simulated or real loads, equivalent to a resistive load greater than 7 kW. The inverter input voltage will be no more than 210 Vdc during the test. Testing will be performed by simulating a loss of input voltage to each 24-hour battery charger. Testing of each as-built 24-hour battery charger will be performed by applying a simulated or real load, or a combination of simulated or real loads. Testing of each 72-hour as-built battery charger will be performed by applying a simulated or real

No. ITAAC No. Design Commitment Inspections, Tests, Acceptance Criteria Analysis load, or a combination of simulated or real loads. Testing of each as-built regulating transformer will be performed by applying a simulated or real load, or a combination of simulated or real loads, equivalent to a resistive load greater than 30 kW when powered from the 480 V MCC. Inspection will be performed for retrievability of the safety-related displays in the MCR.

Inspection will be performed for retrievability of the displays identified in Table 2.6.3-2 in the MCR.

668 C.2.6.09.08a 8.a) Penetrations Inspections will be Penetrations and through the protected performed of openings through area barrier are penetrations through the protected area secured and monitored. the protected area barrier are secured 8.b) Unattended barrier. Inspections and monitored.

openings (such as will be performed of Unattended underground pathways) unattended openings openings (such as that intersect the that intersect the underground protected area protected area pathways) that boundary or vital area boundary or vital area intersect the boundary will be boundary. protected area protected by a physical boundary or vital barrier and monitored area boundary are by intrusion detection protected by a equipment or provided physical barrier and surveillance at a monitored by frequency sufficient to intrusion detection detect exploitation. equipment or provided surveillance at a frequency sufficient to detect exploitation.