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. Y ox:

On September 30, 2021, the U.S. Nuclear R egulatory C ommission (NRC) completed an inspection at Vogtle Electric Generating Plant, U nits 3 and 4. The enclose d inspecti on report documents the inspection results, which the i nspectors di scussed on October 18, 2021 with Mr.

Glen Ch ick, Vogtle 3 & 4 E xecutive Vice President, and other l icensee and co ntractor staff members.

The inspection examined a sample of co nstruction act ivities co nducted und er yo ur C ombined License (COL) as i t relates t o sa fety and compliance wit h the Commissions r ules and regulations and w ith t he conditions of t hese documents. The i nspectors reviewed selected procedures and records, observed activities, and i nterviewed personnel.

NRC i nspectors docu mented five findings of ve ry low sa fety significance (Green) in this r eport.

All of t hese findings i nvolved a violation of NRC requirements. The N RC i s t reating these violations as noncited violations ( NCVs) consistent with S ection 2.3.2.a of t he Enforcement Policy.

In accordance w ith 10 C FR 2. 390 of the NRC' s "Rules of Practice," a co py o f this le tter, its enclosure, and your response ( if any), will be made available electronically for publ ic i nspection in the NRC P ublic D ocument R oom or f rom t he Publicly A vailable Records ( PARS) co mponent of NRC's document s ystem A DAMS. ADAMS i s accessible from t he N RC W eb site at http://www.nrc.gov/reading-rm/adams.html (the Public E lectronic R eading Room).

Should you have any questions c oncerning this letter, pleas e 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. EvenJ. EargleJ. ParentB. Davis DATE 11/11/2021 11/11/021 11/11/2021 11/12/2021 OFFICIAL RECORD COPY U.S. NUCLEAR RE GULATORY CO MMISSION 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

O F FINDINGS

Inspection R eport (IR) 0 5200025/2021005, 05200026/2021005; 07/ 01/2021 through 09/30/2021; Vogtl e Unit 3 & 4 COL, initial t est program and operational p rograms i ntegrated inspection report.

This r eport covers a three -month period of announce d inspecti on Inspections, Tests, Analysis, and Inspection Criteria (ITAAC), p reoperational t est program, start up test program, and operational pr ogram i nspections by r esident and r egional i nspectors. Five findings were determined to be of ve ry low sa fety si gnificance (Green) by t he i nspectors. The significance of most findings i s indicated by their color (Green, White, Y ellow, or R ed) u sing Inspection Manual Chapter ( IMC) 2519, Construction Significance Determination Process. Cross -cutting aspects are determined usi ng IMC 0613, A ppendix F, Construction Cross-Cutting Areas and A spects.

All violations of N RC r equirements a re dispositioned in accordance with t he NRCs E nforcement Policy and the t emporary enf orcement guidance outlined in enforcement guidance memorandum 11 -006. The NRCs pr ogram for ov erseeing the sa fe construction of commercial nuclear pow er r eactors i s desc ribed in IMC 2506, Construction Reactor Oversight Process General G uidance and Basis Document.

A. NRC-Identified and Self R evealed Findings

(Green) A self-revealed constructi on finding of ve ry l ow sa fety significance with an asso ciated noncited vi olation ( NCV) o f Title 1 0 of t he Code of Feder al R egulations ( 10 C FR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and D rawings, w as i dentified for t he licensees f ailure to acco mplish activities i n procedure 3-IDSC-SOP-002, Class 1E A C S ystem

- Division C, V ersion 0 i n accordance with procedur e NMP -AP- 003, Procedure and W ork Instruction U se and A dherence, V ersion 6.1. The licensee enter ed this issue into its corrective action program ( CAP) as condition report ( CR) 50103788, C R 50103789, and CR 50103907.

The performance deficiency was de termined to be of m ore t han minor significance and thus a finding because it represented a su bstantive failure to i mplement an adequ ate program, process, procedure, or q uality ove rsight function. T he inspectors det ermined the f inding was of very l ow sa fety significance (Green) be cause t he event i mpaired only one train of t he IDS. T he inspectors det ermined t he finding had a cross -cutting aspect of H.2, Fi eld Presence, in the area of H uman Performance. Specifically, senior managers di d not ensure t hat operators per forming 3-IDS-SOP-002 had adequate supervisory and m anagement ove rsight o f work ac tivities w hich led to plant equi pment being damaged. [H.2] ( Section 1P01)

(Green) T he NRC i nspectors i dentified a per formance deficiency and construction finding of very l ow sa fety significance and an asso ciated Severity Leve l (SL) I V non -cited violation (NCV) of 10 C FR 50. 59, Changes, tests, and experiments, for t he l icensees f ailure to perform a written sa fety e valuation prior to implementing a change to re move ASME QME 2007 commitments from t he U FSAR in accordance with procedure ND-LI-VNP-002. The licensee entered t his i ssue i nto i ts corrective action pr ogram as C R 50098695.

The inspectors de termined that t he failur e t o follow pr ocedure ND-LI-VNP-002 to per form a 10 CFR 50. 59 evaluation for t he r emoval of ASM E QME 2007 commitments from t he desi gn bases w as a vi olation of 10 CFR 50.59(d)(1), and a performance deficiency. Per t he gui dance in IMC 0613, Appendix E, Examples of Minor Construc tion Issues, the performance deficiency was determined t o be more than minor, and thus a finding, because it represented a su bstantive failure to es tablish or i mplement an adequa te program, process, procedure, or qual ity ove rsight function. Specifically, the removal of Q ME-1-2007 for dyn amic restraints constituted a m ore than minimal i ncrease in the l ikelihood of a malfunction of a structure, sys tem, or component important t o safety and there was a reasonable likelihood that t he ch ange w ould have required Commission review and approval prior t o implementati on 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 P erformance. Specifically, the licensee failed t o ensure t he process for reviewing the change to remove QME-1-2007 for dynamic r estraints was s ystematic in that al l as pects of QME-1-2007 were evaluated for their impact on dynamic restraints. [H.3]

(Section 3P01)

(Green) N RC i nspectors identified a performance deficiency and ITAAC finding of ve ry low safety si gnificance and an NCV of 10 CFR 50.49(e)(5), Environmental qu alification of el ectric equipment i mportant t o s afety f or nu clear pow er p lants, f or the l icensees f ailure to establish the qualified lives of al l nonmetallic co mponents necessary for the completion of t he sa fety f unction of the squib and main steam i solation (MSI) va lves i n accordance with pr ocedure APP-GW-G1-002, " AP1000 Equipment Q ualification Methodology." The l icensee entered this i ssue i nto i ts CAP as CR 50096894.

The performance deficiency was de termined to be mor e t han minor, and t hus a finding, because it is material t o t he acce ptance criteria of t he I TAAC, and the performance deficiency prevented the licensee from meeting the I TAAC D esign Commitment. The i nspectors de termined this finding was of ve ry l ow s afety significance because the f inding did not i mpair a design f unction of t he valves. The inspectors det ermined no cross-cutting aspect applied because the performance deficiency did not reflect cu rrent licensee performance. (Section 3P01)

(Green) N RC inspectors identified an ITAAC f inding and associated NCV of 10 C FR 50. 49.e(5) for t he f ailure to est ablish the qual ified life of t he Containment H igh Range Monitor ( CHRM) d oor gasket in acco rdance with Institute of Electrical and Electronics Engineers S tandard ( IEEE) 323-1974. The licensee enter ed this i ssue into its CAP as C R 50096896 and CAP-IR-2021-6870 and is w orking on corrective actions to address the issue.

The performance deficiency was de termined to be mor e t han minor, and t hus a finding, because it represented an adve rse condition that r endered t he qual ity o f a sys tem, structure, or component ( SSC), unacceptable or i ndeterminate, and r equired su bstantive corrective action. The i nspectors d etermined that the f inding was of very low sa fety significance because the finding was no t associated with a sys tem o r s tructure i n risk importance table. The inspectors det ermined t he finding was a ssociated w ith the cross -cutting aspect of Consistent Process in the area of H uman Performance. Specifically, the l icensee failed to u se a consistent, systematic a pproach t o make decisions w hether t he AP1000 harsh environment included wetted environments. [H.13] (Section 3P01)

(Green) N RC 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 acc ordance with IEEE 317, Section 6.2.10. Specifically, the licensee failed to test or analyze the as installed as sembly including cable configurations, terminations, termination box, raceways, and their res pective interactions. The licensee entered this is sue into its CAP as CR 50104476 and is working on corrective actions to address the issue The performance deficiency was de termined to be mor e t han minor, and t hus a finding, because it represented an adve rse condition that r endered t he qual ity o f an SSC un acceptable or indeterminate and r equired substantive corrective action. T he inspectors det ermined the f inding was of very l ow sa fety si gnificance because the l icensee was able t o sh ow t hrough addi tional analysis t hat t he finding did not i mpact a design function of the EPAs. The inspectors determined the finding was asso ciated w ith the cr oss-cutting aspect of Change Management in the area of Human Performance. Specifically, the licensee leadership failed to use a systematic pr ocess for e valuating and implementing changes to as i nstalled SSCs compared t o configurations that wer e environmentally qual ified so that nuclear sa fety r emained the ove rriding priority. [H.3] (Secti on 3P01)

B. Licensee-Identified Violations

None

REPORT DE TAILS

Summary of P lant C onstruction Status

During this report pe riod for U nit 3, t he license e complet ed various a ctivities t o sa tisfy asp ects o f the Vogtle Unit 3 oper ational pr ograms and initial t est program. The l icensee completed hot functional testing act ivities w hich included testi ng the reactor co olant s ystem, residual hea t removal system, pas sive core co oling system, main steam sy stem, etc. at elevated temperatures and pr essures. The l icensee performed post hot functional t esting inspection of the reactor i nternals to ensur e that t here was no observable damage or l oose parts. The licensee performed pr eoperational and component tests o f various S SCs and their co ntrol systems, e.g. Class 1E DC and UPS sys tem (IDS), protection and monitoring system (PMS),

and plant co ntrol sys tem. Additionally, the l icensee performed testing of ai r oper ated valves and motor ope rated va lves; and performed response time testing of per manently i nstalled plant instrumentation.

During this report pe riod for U nit 4, t he license e continued with integrat ed flush acti vities b y flushing portions of ch emical and volume co ntrol s ystem sp ent fuel pool s ystem ( SFS), and RNS. The l icensee completed the i nitial pum p r uns for t he component cooli ng water s ystem and service water sy stem, and the initial co mpressor r un for t he c ompressed and instrument ai r system. In addition to co ntinuing with the i nstallation of pl ant SSCs, the licensee is making preps for t he upco ming secondary h ydrostatic t est.

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 inspecti on 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 Significanc e 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, Ver sion (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 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 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 2007, Section QDR, Qualification of Dynamic Restraints, to ASME B oiler and P ressure 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 appropr iate 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 informa tion 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 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 signifi cant 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 asso ciated w ith a repetitive, N RC-identified omission of a pr ogram c ritical attribute.

The inspectors de termined the f inding was a performance deficiency with a significance very l ow sa fety significance, GREEN, bec ause the finding was no t associated with a system o r s tructure in risk i mportance table.

The inspectors de termined the f inding was i ndicative of pr esent l icensee performance and was asso ciated with the cross -cutting aspect of Consistent P rocess in the ar ea of Human Performance, in accordance with IMC 06 13, A ppendix F, Construction Cross -

Cutting Areas and A spects, da ted November 4, 2020. Specific ally, the licensee failed to use a consistent, sys tematic appr oach to m ake decisions w hether t he AP1000 harsh environment included wetted env ironments. [H.13].

Enforcement

10 CFR P art 50.49.e(5), Aging, r equired, in part, that equipment qual ified by test must be preconditioned by nat ural or ar tificial ( accelerated) agi ng to its en d-of-installed life condition. C onsideration must be gi ven to all s ignificant t ypes o f degradation which can have an e ffect on t he functional ca pability o f the equipment.

Contrary t o the above, since October 2020, the licensee fail ed to precondition equipment qual ified by test by nat ural o r a rtificial ( accelerated) agi ng to its end-of-installed life condition and consider al l significant types of degradation which c an have an effect on the f unctional ca pability of t he equi pment. Specifically, the licensee failed to consider the known po tential f ailure m odes (aging) of the CHRM door g asket to determine its qual ified life which can have an ef fect on t he f unctional ca pability of the CHRM.

The licensee entered this i ssue into i ts CAP as C R 50096896 and C AP-IR-2021-6870 and is w orking on corrective actions to address the issue. This i ssue does not represent an i mmediate safety co ncern because there was no fuel i n t he reactor and these components were not r equired to perform t heir safe ty function at the time of t he inspection.

Because this vi olation was not r epetitive or w illful, and w as en tered i nto the licensees corrective action program, this violation is bei ng treat ed as a NCV consistent w ith Section 2.3.2.a of the NRC E nforcement P olicy (NCV 05200025/ 2021005-04, Failure to Demonstrate Q ualification of C ontainment High Radiation Monitor D oor G asket).

NCV 05200025/ 2021005-05, Fai lure t o Q ualify EPAs i n Accordance with IEEE 317

Introduction

NRC i dentified a Green I TAAC f inding and associated violation of 10 C FR 50, A ppendix B, Criteri on III, f or t he failur e to seismically or environmentally qual ify the EPAs i n accordance with IEEE 3 17, S ection 6. Specifically, the l icensee failed to t est or analyze the as i nstalled assembly i ncluding cable configurations, terminations, termination box, raceways, and their r espective 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, S ubsection 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 f ire 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 f ire 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 as 05200025/2020009- 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 Requirements

05200025/2021005- 03 NCV Open Failure to Demonstrate Qualification of Valve Nonmetallic Parts

05200024/2021005- 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 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 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 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 B attery 1B onl y), Ver. 7 APP-DB01-Z0- 001, " Design Specification for C lass 1E 250 V DC B atteries and Racks," R ev. 10 3-IDS-ITPP-501, " Class 1E D C and UPS P reoperational Test ing," V er. 5.1 CR 50080247 Section 3P0 1

APP-GW-G0X-003, AP1000 Commodity Loca tor Codes, R ev. 11 APP-GW-G1X-001, Governing AP1000 Design Codes and Standards, Rev. 10 APP-GW-VP- 010, Equipment Q ualification Methodology and Documentation Requirements for AP1000 Safety-Related Valves and Valve Appurtenances, Rev. 3 CR 50098695 LDCR-2016-107, Snubber Q ualification Changes, Rev. 2 LDCR-2016-107, Snubber Q ualification Changes, Rev. 1 APP-GW-VPD-002, AP1000 Safety R elated C omponent Li st, Rev. 11 APP-GW-G1-002, AP1000 Plant E quipment Qualification Methodology, Rev. 3 APP-GW-N2C- 009, Beta R eduction Factors and Gamma E quivalency for Various S hielding Materials, Rev. 0 APP-PV70-T5- 001, Qualification Plan for S afety-Related Squib Valve Actuators, Electrical Connector A ssemblies, and Bracket Assemblies for W estinghouse Electric C ompany f or use in Westinghouse AP1000 Nuclear P ower P lants, Rev. 5 LTR-EQ-18-23, Activation Energy f or TE C onnectivity Splice Materials, 05/22/2018 ND-AD-002, Nuclear D evelopment C orrective Action Program, Rev. 29.1 ND-EN-VNP-006, Equipment Q ualification Data Package (EQDP) R eview and Acceptance, Rev. 4.0 ND-LI-VNP-002, Applicability Determination and 50. 59 / D eparture Screening for V EGP 3& 4, Rev. 18 ND-LI-VNP-002-F06, 50.59 / D eparture Screening for V ogtle 3&4, Rev. 6 ND-LI-VNP-003, 50.59 / D eparture Evaluations for Vogtle 3&4, Rev. 8.0 ND-LI-VNP-007, Licensing Document Change R equests for V EGP 3& 4, Rev. 10 NMP-AD-008, Applicability D eterminations, Rev. 21.1 SV0-EY20-VPC-001, Vogtle Units 3 and 4 EY 20 Cale Splice Thermal A ging and Post-DBA Aging Calculation, Rev. 0 APP-PV64-Z0- 001, Desi gn Specification Mai n Steam I solation, ASM E BPVC,Section III, C lass 2, Rev. 7 APP-PV70-Z0- 001, Squib (Pyrotechnic Actuated) V alves, ASME B oiler an d Pressure Vessel Code Section III, C lass 1, Rev. 7 App-SS30-Z0- 002, Design and Fabrication Requirements for H ydraulic S hock Absorbers (Snubbers), Rev. 4 APP-EY20-VBR-002, Equipment Qualification Data Package f or C lass 1E Low V oltage Cable Splices f or U se in the AP1000 Plant, Rev. 0 APP-PV64-VBR-012, Equipment Qualification Data Package f or Main Steam Isolation Valves for U se in the AP1000 P lant, R ev. 0 APP-PV70-VBR-005, Equipment Qualification Data Package f or 14 S quib V alves for U se in the AP1000 Plant, Rev. 1 SV3-1050- P2-0014, Nuclear Island Containment and Shield BLDG General A rrangement Operating Deck EL 135'-3, R ev. 5 SV3-PMS-VPR-009, Mild Environment Abnormal Tem perature and R elative Humidity E xtremes Report for N IS and RMS Cable Connection Assemblies f or U se in the A P1000 Plant, Rev. 0 SV3-EJ01-VBR-002, E quipment Qualification Data Package f or E J01 C lass 1E Ju nction Boxes for U se in the AP1000 P lant, Rev. 0 SV3-EY20-VBR-002, L ow V oltage Cable Splices, Rev. 0 SV3-1050- P2-0014 N uclear Island Containment and Shield BLDG General A rrangement Operating Deck EL 135'-3, Rev. 5

SV3-PMS-VPR-009 M ild Environment Abnormal Temperature and Relative Humidity E xtremes Report for N IS and RMS Cable Connection Assemblies f or U se in the A P1000 Plant, Rev. 0 SV3-RMS-V0-001 A P1000 Radiation Monitoring System Vendor O utline Drawing Package, Rev. 1 SV3-RMS-VBR-002 E quipment Q ualification Data P ackage for t he R adiation Monitoring System (RMS) C ontainment High Range Monitor ( CHRM) and Fi eld Cable Connections a t the EPA Feedthroughs, Rev. 0 SV3-RMS-VBR-007 C ontainment H igh Range Monitor D esign Basis A ccident Tes t 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 P re-Plans, Ver. 3.0 B-PFP-ENG-001-F3110, A ux. Bldg. non -RCA, El 66-6 B-PFP-ENG-001-F3114, A ux. Bldg. non -RCA, El 117-6 B-PFP-ENG-001-F3128, A ux. Bldg. & P CS V alve Room B-GEN-ENG-008, Fire E quipment Fun ctionality a nd Fire Protection Impairments ( FPI)

Requirements, Version 1.0 B-GEN-ENG-008-GL02, Fire Equipment Func tionality and Fire P rotection I mpairments ( FPI)

Basis, Version 1. 0 ND-LI-NP-007, Licensing D ocument Change R equests for V EGP U nits 3 & 4, V ersion 9.0 NMP-ES- 035- 005, Fi re Protection Alternative Compensatory Measures, Version 6.0 NMP-ES- 035- 006, Fi re Protection Program I mpact Screen and D etailed 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 C hange Evaluation, V ersion 4.0 NMP-ES- 095- 001, D ocumentation Change Processes, Version 1.0

Work Orders SNC921209, FSP Individual Fire Detector Test Package SV3-1230- CCW-CV2443, Aux B uilding Battery R ack Walls (32, 33, 34, 36, & 37) up to El. 100-0, cl osed 10/ 29/2020 SV3-1230- CCW-800001, 1084 to 115 -6, CL W all, Concrete P lacement & P ost-Placement (Wall 89), closed 2/ 7/2019

Miscellaneous 3-AOP-902-B, Background Information for 3 -AOP-902, Fi re Response E mergency, R ev. G.6 APP-IDS-E8-001, Class 1E D C and U PS S ystem S pecification Document, R ev. 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 P rotection Regulation Structure (Applicable NFPA C odes of R ecord), Revised 6/25/2021 FPRA Calculation List, R evised 6/25/2021 Maximo FP S P M Li st, Revised 6/25/2021 Fire Pre-Plan I ndex and Status Li st, Revised 6/25/2021 SV3-1200- ITR-800797, Fire Barrier Inspection Report, U nit 3: ITAAC 3.3.00.07c.ii.a, 3.3.00.07c.ii.b, Rev. 0 VS3-ITAAC-ST-2.3.04.10, FP S Individual Fi re Detectors - ITAAC: SV3-2.3.04.10, Rev. 0 WCAP-15871, A P1000 Assessment Against NFPA 804, R ev. 1

Corrective Action D ocuments R eviewed CR 50092969, E quivalent Composite Firewall material not pa rt of a l isted assembly, 5/18/2021 CR 50004139, C able Route Design Discrepancy with COL A ppendix C, 8/31/2018 CAR 50007691, C able Route Design Discrepancy with COL Appendix C, 8/31/2018 TE 60003574, C able Route Design Discrepancy with COL A ppendix C, 8/31/2018

Condition Reports Generated During I nspection CR 50103154, W CAP-15871 incorrect statement on listing of fire walls an d 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 Functional Qualification Report for 14 Squib Valves Installed in AP1000 Plants, Rev. 0

APP-PV70-VPR-014, ASME Q ME-1 Functional Qualification Report for 8 S quib Valves Installed in AP1000 P lants, Rev. 0 APP-PV70-VPR-016, QME-1 Test ing of a 14 -2500# Squib Valve for W estinghouse Electric Company, LLC Jo b N umber T57622 ( PR028955), Rev. 0 APP-PV70-VPR-021, Test Report of QME Te sting Performed on 8 inch HP S quib Valve, Re. 0 APP-PV70-VQQ-002, 14 A DS S quib Valve Assembly D ata P ackage, 08/2018 APP-PV70-VQQ-053, Lot A cceptance Test ing, Rev. 0 APP-PXS-PLR- 010, AP1000 Direct Vessel Injection Line A ( APP-PXS-PLR-010) P iping Stress Analysis R eport, Rev. 6 APP-PXS-PLR- 020, AP1000 Direct Vessel Injection Line B ( APP-PXS-PLR-020) P iping Stress Analysis R eport, Rev. 6 APP-PXS-PLR- 030, Piping Analysis R eport for L oop 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 2020), Vogtle 3&4 Application Report for T ricentric N PS 10 Motor-Operated B utterfly Valves

Procedure NMP-ES- 013- 007, I ST Program D ata Trending and Evaluation, R evision 3.0

Section 3P06

Drawing APP-MI01-V2-133, Reactor i nternals co re bar rel a ssembly, Rev. 3

Specifications APP-MI01-Z0- 001, AP1000 Reactor i nternals functional specification, Rev. 5 APP-MI01-Z0- 101, AP1000 Reactor vessel i nternals desi gn specification, Rev. 10 APP-MI01-Z0- 600, AP1000 Reactor vessel i nternals f abrication requirements, Rev. 3

QA document QR 3343, Q uality r elease and certificate of conformance, R ev. 0

4. OTHER I NSPECTION R ESULTS

Section 4OA 5

.01

B-GEN-ITPCE-008, " Class 1E and N on-Class 1E Battery Capacity Test ing," R ev. 7 APP-DB01-Z0- 001, " Design Specification for C lass 1E 250 V DC B atteries and Racks," R ev. 10 3-IDS-ITPP-501, " Class 1E D C and UPS P reoperational Test ing," V er. 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 offlow 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 no load RCS pressure and temperature conditions will be performed. Analyses will be performed to convert the measured pre-fuel load flow to post fuel load flow with 10 percent 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 befunctions. 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. RCS V001A/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 internalsi) Not used peri) Not used per will withstand the Amendment No. 151. Amendment No.

effects of flow induced ii) A pre-test151. 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 providesi) A containmenti) 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 operate dvalves 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 ransfer the test will be above reeasurn the top of the HX. The tt test will continue unt iladjto account the hot leg f differencn t temperature is ei and fotp T 250°F. trur the number ofubes

603 2.6.03.04c 4.c) Each IDS 24-hour Testing of each 24-The battery terminal battery bank supplies a hour as built 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-hourhour 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 chargera 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 l oadof 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 t hedesign 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-builtidentified 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.