ML033020041

E-mail from C. Ogle to P. Fillion and T. Mckenzie Regarding Comments on McGuire IR-03-07

ML033020041
Person / Time
Site:	McGuire, Mcguire
Issue date:	07/08/2003
From:	Ogle C NRC/RGN-II
To:	Fillion P, Mckenzie T NRC/RGN-II
References
FOIA/PA-2003-0358 IR-03-007
Download: ML033020041 (44)

Text

ECharles R. Ogle -

MY COMMENTS ON MCG IR 03-07 From:

Charles R. Ogle To:

Fillion, Paul; Thomas, McKenzie Date:

7/8/03 11:06AM

Subject:

MY COMMENTS ON MCG IR 03-07 Overall, I thought this was a good report. Attached are some of my comments. I've also attached a comparison version so that you can see what changes were made between what you gave Charlie and what I signed. (Needs to be opened in WP to see the changes.)

WHile the attacment identifies areas for Improvement on the report, it does not diminish the fact that I think that the inspection and issue raised were outstanding.

CC:

Payne, Charlie Page 1 I/3

, Charles R. Ogle - compare mcg03O07.wpd Page 1 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I Xt SSAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET SW SUITE 23T85 ATLANTA, GEORGIA 30303-8931 1

July XX3, 2003 Duke Energy Corporation ATTN: Mr. DG. JammlPeterson Vice President McGuire Nuclear Station 12700 Hagers Ferry Road Huntersville, NC 28078-8985

SUBJECT:

MCGUIRE NUCLEAR STATION - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 50-369/03-07 AND 50-370/03-07

Dear Mr. jamiiPeterson:

On May 23, 2003, themU.S. Nuclear Regulatory Commission (NRC) completed an inspection at your McGuire Nuclear Station, Units 1 and 2. The onclood rport documents the inspection findings which were dicussodAn interim exit was held with Mr. D. Jamil and other members of your staff on May 22, 2003, to discuss the results of that effort. Following completion of additional review in the Region II office, a final exit was held with you and other members of your staff on July 2, 2003. The enclosed report documents our findings from this inspection.

The inspection examined activities conducted under your licenses as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your licenses. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents twethree findings that a

ibi*ehave potential safety significance greater than very low significance, however, a safety significance determination has not been completed. OpeThese findings did not present an immediate safety concern afdat the time of the interim exit. However, your subsequent analyses of one of the findings associated with Fire Area 16/18 resulted in identification of additional cables associated with reactor protection system instrumentation (and possibly other equipment) required for safe shutdown located in the same fire area that could be susceptible to fire damage. Upon discovery of this condition on June 10, 2003, a fire watch was put in place on Juno 10, 2003,established as a compensatory measure.

In addition, tho roport documents ono NRC identified finding which was dotorminod to involvo a violation of NRC roquiromonts. Howovor, the ignificance of this finding has not boon dotorminod. Also, one licenseo identified violation is listed in this report. If you contest any violation in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the United States Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory

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Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the McGuire facility.

In accordance with 10 CFR 2.790 of the NRC's Rules of Practice," a copy of this letter and its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of-

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NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.pov/readina-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA Charles R. Ogle, Chief, Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-369, 50-370 License Nos.: NPF-9, NPF-17

Enclosure:

Inspection Report 50-369, 370/03-07 w/

Attachment:

Supplemental Information cc w/encl:

C. J. Thomas Regulatory Compliance Manager (MNS)

Duke Energy Corporation Electronic Mail Distribution M. T. Cash, Manager NucleaF Regulatory LieRniRgIssues & Affairs Duke Energy Corporation 526 S. Church Street Charlotte, NC 28201-0006 Lisa Vaughan Legal Department (PB05EEC11X)

Duke Energy Corporation 422 South Church Street Charlotte, NC 28242 Anne Cottingham Winston and Strawn Electronic Mail Distribution Beverly Hall, Acting Director Division of Radiation Protection N. C. Department of Environmental Health & Natural Resources Electronic Mail Distribution

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DEC (cc w/encl cont'd - See page 3)

(cc w/encl cont'd)

County Manager of Mecklenburg County 720 East Fourth Street Charlotte, NC 28202 Peggy Force Assistant Attorney General N. C. Department of Justice Electronic Mail Distribution 4

Distribution w/encl:

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lOFFICE R:DRS I RI:DR IR:DRS I RI:DRS RI:Consultant RIl:DRS TRII:DRP ISIGNATURE I RA I RA I RA I RA I RA I RA I RA

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. -'f-6 U.S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos.:

License Nos.:

Report Nos.:

Licensee:

Facility:

Location:

Dates:

Inspectors:

Approved by:

50-369, 50-370 NPF-9, NPF-17 50-369/03-07 and 50-370/03-07 Duke Energy Corporation McGuire Nuclear Station, Unts-awnd4 12700 Hagers Ferry Road Huntersville, NC 28078 May 5 - 9, 2003 (Week 1)

May 19 - 23, 2003 (Week 2)

P. Fillion, Reactor Inspector R. Maxey, Reactor Inspector B. Melly, Fire Protection Engineer (Consultant)

R. Schin, Senior Reactor Inspector (April 14-17, 2003)

M. Thomas, Senior Reactor Inspector (Lead Inspector)

Charles R. Ogle, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

Charles R. Ogle - comparemcg 03_07.wpd Page 7 7

SUMMARY

OF FINDINGS IR05000369/03-07, R05000370103-07; Duke Energy Corporation; 61/9-23/20305/05-09/2003 and 05/19-23/2003; McGuire Nuclear Station, Units 1 and 2; Triennial Fire Protection The report covered a two-week period of inspection by regional inspectors and a consultant.

Three unresolved items with potential safety significance greater than Green were identified.

The significance of most findings Is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG 1649, Reactor Oversight Process," Revision 3, dated July 2000.

A.

Inspector-NRC-dentified and Self-Revealing Findings Cornerstone: Mitigating Systems TBD? The team identified a violation !RV4weM;gin that Train A and Train B cables associated with theredundant reactor protection system instrumentation (and possibly other equipment) important to safe shutdown were located in the same fire area (Fire Area 16/18) and were not protected from fire damage, as required by McGuire's fire protection program.-

This finding is unresolved pending determination of the systems affected and completion of a significance determination. Theis finding is greater than minor because it was associated with the equipment performance attribute and affected the objective of the mitigating systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events in that instrumentation important for post-fire safe shutdown weukdcould be lost. Tho finding roprosontod an operability conom, which tho liconsoo resolved by posting a fire watch in tho aroa When assessed in combination with the finding related to inadequate protection of auxiliary feedwater system cables and equipment required for safe shutdown In Fire Area 16/18 (also discussed In this inspection report), this finding may have potential safety significance greater than very low significance. (Section 1 R05.03.b.1)

TBD-The team identified a violation in that the turbine driven auxiliary feedwater (TDAFW) pump suction supply valve 2CA0007A was not evaluated in the licensee's Fire-Protection Program (i.e., safe shutdown analysis} for potential impact on safe shutdown in the event of a fire where the TDAFW} pump is required for safe shutdown. The valve could spuriously eeseoperate due to fire damage and adversely affect the TDAFW pump.

The finding is unresolved pending completion of a significance determination. The finding is greater than minor because spurious closure of the valve could damageit was associated with the equipment performance attribute and affected the objective of the mitigating systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events. This finding may have potential safety

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significance greater than very low significance because the standby shutdown system relies on the TDAFW pump for decay heat removal, and veriously degrade the decay heat removal function would be seriously degraded if the TDAFW pump were damaged due to closure of valve 2CA0007A. (Section 1 R05.04.b.2)

B.

Licensee-Identified Violations TBD-The physical protection of cables and equipment relied upon for safe shutdown (SSD) of Unit 2 during a fire in the Train A Switchgoar Room!EloctricaIElectrIcal Penetration Room (Fire Area 16/18) was not adequate. Train B electrical cables, associated with the 2B motor driven auxiliary feedwater pump discharge valve 2CA0042B to steam generator 2D, were located in the Train A Electrical Penetration Room (Fire Area 16/18) without adequate spatial separation or fire barriers as required by the McGuire fire protection program. Local, manual operator actions (which had not been reviewed and approved by NRC) would be used to achieve and maintain SSD of Unit 2 in lieu of providing adequate physical protection for the electrical cables associated with valve 2CA0042B.

This finding is unresolved pending completion of a significance determination. The finding is greater than minor because it was associated with the 2B motor drivon auxiliary feedwator pump dirchargo valve 2CA0042B to steam gnorator Q1, were-lerated !R Train A Eloctrical Ponotration Room (Firo Aroa 16/18) without adequate spatial reparation or firo barriors a rjquired by the Firo Protoction Program. Local, m~nltnl

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ucod to achiove and maintain SSD of Unit 2 in lieu of-providing adequate physical-protoction for the electrical cables acsociated with valve 2CAGO42B.

This finding i unresolved pending completion of a 6ignficanGe determination. Tho finding i greator than minor bocausooquipment performance attribute and affected the objective of the mitigating systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events in that fire damage to the unprotected cables could prevent operation of SSD equipment from the main control room aRd boaurA it affonts the msigating SyStemnr coFRnrtrono objective. When assessed in combination with the inadequate reactor protection system cable separation finding (also discussed in this inspection report), this finding may have potential safety significance greater than very low significance. (Section 1 R05.03.b.2)

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Report Details

1.

REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity 1 R05 FIRE PROTECTION O.,pnmp I;n, oirnd i Ahivn.and Maintain PO i

SA i

Rhutdown Fire Protection The purpose of this inspection was to review the McGuire Nuclear Station (MNS) fire protection program (FPP) for selected risk-significant fire areas. Emphasis was placed on verification that the post-fire safe shutdown (SSD) capability and the fire protection features provided for ensuring that at least one redundant train of safe shutdown systems is maintained free of fire damage. The inspection was performed in accordance with the Nuclear Regulatory Commission (NRC) Reactor Oversight Program using a risk-informed approach for selecting the fire areas and attributes to be inspected. The team used the licensee's Individual Plant Examination for External Events (IPEEE) and performed in-plant walk downs to choose four risk-significant fire areas for detailed inspection and review. The four fire areas selected were:

Fire Area 4, Auxiliary Building (AB) Common Area; AB +716 feet elevation Fire Area 13, Battery Rooms; AB +733 feet elevation common area Fire Area 16/18, Unit 2 Train A Electrical Penetration Room/2ETA 4160 volt Switchgear Room; AB +750 feet elevation Fire Area 24, Main Control Room (MCR); AB +767 feet elevation For each of the selected fire areas, the team focused the inspection on the fire protection features, and on the systems and equipment necessary for the licensee to achieve and maintain safe shutdown conditions in the event of a fire in those fire areas.

The team evaluated the licensee's firo pFt8ctio Pogram (FPP} against applicable requirements, including Operating License Conditions 2.C.4 and 2.C.7, Fire Protection Program, for Units 1 and 2, respectively; Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), Appendix R, Sections I I. G, J, L, and 0; 10 CFR 50.48; Appendix A to Branch Technical Position (AP)-Auxiliary and Power Conversion Systems Branch-(APG813) 9.5-1, Guideline for Fire Protection for Nuclear Power Plants; related NRC Safety Evaluation Reports (SERs); McGuiro Nuclear Station (MNS) Updated Final Safety Analysis Report (UFSAR), Section 9.5.1; UFSAR Section 16.9, Selected Licensee Commitments (SLC); and plant Technical Specifications (TS). The team evaluated all areas of this inspection, as documented below, against these requirements.

Th tam rosvi d tho lIconcees Indit;iduatl lanrt Examinaton for External Esont IPSE -and -4

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C harles R. Ogle - compare-mcg_03_07.wpd Page 10 10 and roviow. Tho four fire aroas oloctod wore:

Fire Area 4: Auxiliary Building Common Aroa a firo in this area would involvo altom;it shutdon fFrom tho staRdby rhutdown facility (SSF) using tho taRdby shutdown system (SSS)

Fire Area 13: Battery Rooms Common Area a firo in this aroa would involve altomativo shutdow fmA the SSF uring the SSS Fire Area 16118: Unit 2 Train A 4160 Volt Switchgoar Room/Eloctrical Ponotration Room a firo in this aroa would involve shutdown from tho main cOtr oom using TraiR1n euipment Fire Aea 24: Main Control Room (MCR) a firo in thi aroa would involvo altornative shutdown from tho SSF using tho SSS

.01 Systems Required to Achieve and Maintain Post-Fire Safe Shutdown

a.

InsDection Scove The team reviewed the licensee's FPP documontoddescribed In UFSAR Section 9.5.1; the MNS Fire Protection Review, safe shutdown analysis (SSA); fire hazards analysis (FHA); safe shutdewR-(SSD} essential equipment list; and system flow diagrams-to identify the components and systems necessary to achieve and maintain afe-shutdownSD conditions. Speif!G liconcoc documnts, aiculation, and drawings.

roviowod during this inRspctine aFO listed in Attachmont 1. The objective of this evaluation was to assure tho SSD eqUipMent and post ie SSD anal#cal app9rac4h woF re nsistent with aRd satisfiod the Appendix R Foa F pooremFance crit9ria for SSD.

For each of the selected fire areas, the team focused on the fire protection features, and on the systems and equipment necessary for the licensee to achieve and maintain SSD in the event of a fire in those fire areas. SThe following Unit 2 systems aid/~and components were selected for reviewicluded: 6 Standby shutdown system (SSS); Unit 2 etandby Standby makeup pump (SMP) 2NVPU0046 aXl-SMP suction supply motor poratod valve(MOV) 2NV842AC;-auxiliy Auxiliary feedwater (AFW) suction supply M4V6valves 2CA007Aj and 2CA009B-2VAI 61 C, ad 2GA 62; actor Reactor coolant pump (RCP) seal water return isolation valve 2NV94ACj-Pressurizer power operated relief valve (PORV) 2NC34A-and-PORV Isolation valves 2NC33A; Unit 2 prosurizer Pressurizer heaters Nos. 28, 55, ai:W-56:ractor Reactor vessel head vent valves 2NC272AC and 2NC273AC; and hating Heating, ventilation, and air conditioning (HVAC)

Specific licensee documents, calculations, and drawings reviewed during this inspection are listed in the attachment.

b.

Findings

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ae1 11 No findings of significance were identified.

.02 Fire Protection of Safe Shutdown Capabilitv

a.

Inspection Scone The team reviewed the fire detection system protecting Fire Areas 4, 13, 16/18 and 24 to assess the adequacy of the design and installation. This was accomplished by reviewing design drawings, ceiling beam location drawings, and National Fire Protection Association (NFPA) 72E (code of record 1974 edition) for detector location requirements. The team reviewed the McGuire Fire Protection Code Deviation Calculation to determine if there were any outstanding code detector deviations for the selected areas. The team walked down the fire detection and alarm systems in Fire Areas 13,-46 and 4816/18 to evaluate the installed detector locations relative to the NFPA 72E location requirements. Additionally, the team reviewed the surveillance test procedures for the detection and alarm systems to determine compliance with-he UFSAR Sections 9.5.1 and 16.9.

v The team reviewed the adequacy of the design and installation of the fire suppression system protecting the nuclear service water (RN) pump area in Fire Area 4. This was accomplished by reviewing the engineering design drawings, suppression system hydraulic calculations, as-built system configuration and NFPA 13 (code of record 1978 edition) for sprinkler system location requirements. The team also reviewed the McGuire Fire Protection Code Deviation Calculation for the RN pump sprinkler system to determine the adequacy of the system to control a fire in this area utilizing the 2-1/2 inch by-pass lines as the sole means of supplying the sprinkler system.

The team reviewed the fire hose stations in Fire Areas 4, 13,167/18 and 24 to assess the adequacy of the design and installation. This was accomplished by reviewing the fire plan drawings, engineering mechanical equipment drawings, pre-fire strategies and NFPA 14 (code of record 1976 edition) for hose station location requirements and effective reach capability. Team members also performed a field walkdown of the selected fire areas to ensure that hose stations were not blocked and to compare hose station location drawings with as-built plant locations.

b.

Findings The team identified an unresolved item (URI) involving the adequacy of the suppression system for Fire Area 4. AltemativeDedicated shutdown (DSD) using the SSS was designated by the licensee for a fire in this area. 10 CFR 50, Appendix R, Section III.G.3 (alternative or dedicated shutdown) requires that fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under consideration.

THowever, the fire suppression systom for Firo Area 4 WaR not installed in accordanco with 10 CFR 50, Appondix R, Soction Il.G.3. The system in Fire Area 4 was a partial automatic sprinkler system offoctivoly protoctingdesigned to protect the RN pumps and the area 20 feet north of these pumps. The area protected by this sprinkler system was located between Gcolumn lines 54-58 and EE-GG. The majority of Fire Area 4 was not provided with automatic sprinkler protection as required by 10 CFR 50, Appendix R,

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12 Section Ill.G.3 for altomativo and dedicated shutdown.

This issue was previously identified by the NRC (URI 50 369/84 28 01, 370/84 25 01) in 1984 during an Appendix R inspection (URI 50-369/84-28-01, 370/84-25-01). The licensee considered this issue to be a potential backfit per 10 CFR 50.109 (letter dated September 4, 1984, from H.B. Tucker, Duke Power Company, to H.R. Denton, NRC Office of Nuclear Reactor Regulation). The URI was roviowed and closed in NRC Inspection report (IR) 50-369,-370/87-34. The team noted that, subsequent to closure of the URI, licensee Fire Protection Functional Audit SA-99-04(MC)(RA)(FPFA) dated April 9, 1999, identified that MNS did not meet separation and detection/suppression criteria for alternative or dedicated shutdown capability as required by 10 CFR 50, Appendix R, Section III.G.3. During tWethe current inspection, the team questioned whether the previous NRC reviews of the sprinkler system for this fire area included an evaluation of the risk impact associated with not providing adequate sprinkler coverage for the RN cabling in this fire -area. The team informed the licensee that this issue w1Uwould be reviewed ther to determine if the lack of sprinkler coverage in this fire area has an Impact on risk. This iEuo i identified as URI 60 369,370/03 07 01, Firo Suppression System for Altomativo Shutdown Areas not in Accordanco with 10 CFR 50, Appondix R, Soction Il.G.3. The team noted that a similar conditions, rgarding the fixed firo supprossion cystom complying with 10 CFR 50, Appondix R, Soction Il.G.3, was applicablo to other MNScondition exists in other fire areas where al4e.nativededicated shutdown capability using the SSS was designated by the licensee (xamplos include Rirm Arwas 4 ad 1). This 1iv14 i unrF8soled pending furtheF NRC eview uing isks insights to dotormino if a 10 CFR 60.109 (backf it). Pending determination of whether a backfit evaluation is warranted, this issue is identified as URI 50-369, 370/03-07-01, Fire Suppression System for Dedicated Shutdown Areas Not in Accordance with 10 CFR 50, Appendix R, Section III.G.3.

.03 Post-Fire Safe Shutdown Circuit Analysis

a.

Inspection Scope The team reviewed the adequacy of separation and fire barriers provided for the power and control cabling of equipment relied on for SSD during a fire in the selected fire aear/40Fesareas. On a sample basis, the team reviewed the SSA and the electrical schematics for power and control circuits of SSD components, and looked for the potential effects of open circuits, shorts to ground, and hot shorts. This review focused on the cabling of selected components mof the GhagiR§/afetY-ijemien harging/makeup system, reactor coolant system (CS) and AFW system. The team traced the routing of cables by using the cable schedule and conduit and cable tray drawings. Walkdownc wore performed to comparo cables indicated on the drawings with actual plant installation. Circuit and cable routings were reviewed for the following equipment:-

ORN4AC, TuFbin -driv9Turbine Driven AFW Suction Supply Valve 2CA0007A, Turbine-dWveRTurbine Driven AFW Suction Isolation Valve 2CA009B, Moto -driveRMotor Driven AFW Suction Isolation Valve 2CFLT6080, 6090, 6100, 61 10, Steam Generator Level Transmitters 2NCLT5151, Pressurizer Level Transmitter

Charles R. Ogle - comparemcg 03_07.wpd Page 13 13 2NC34A, 33A-Pressurizer PORV amd 2NC33A, PORV Isolation Valve 2NC272AC, 273AC, Reactor Vessel Head Vent Valves 2NVPU0046, Standby Makeup Pump-(SMP4 2NV94AC, RCP Seal Water Return Isolation Valve 2NV842AC, SMP Suction Isolation Valve 2NV1012C, SMP Discharge to Containment Sump Isolation Valve Pressurizer Rheaters Nos. 28, 55, 56 The team also reviewed licensee studies of overcurrent protection an abnthfor alternating current (AG) and direct current4DG* systems to identify whether fire-induced faults could result in defeating the safo shutdownSSD functions.

b.

Findings Findings associated with valves 2CA0007A, 2NC34A, and 2NC33A are discussed in Section 4-R05.04 of this inspoction roport.

X-.

Ron-Aor Pmt R

R.

1.

Inadequate Separation of Cables Associated With Safe Shutdown Instrumentation

==

Introduction:==

A finding with potentially greater than very low safety significance was identified in that redundant instrumentation (and possibly other equipment) important to safe shutdownSSD could have beeo damaged by a fire in Fire Area 16/18. This finding involved a violation of NRC requirements. This finding is an URI pending a determination of the systems affected by the licensee and completion of the significance determination process (SDP).

==

Description:==

Fire Area 16/18 is the Unit 2 Train A electrical penetration room/2ETA 4160 volt (V) switchgear room. and the associated HVAC equipment room 805A. Train B equipment controlled from the nai GF#FGIMCR room was 44ende4ddesignated as to-be usedhe SSD train for a fire in this area according to the aRalysisSSA and plant procedures (i.e., this fire area complies with 10 CFR 50, Appendix R, Section III.G.2).

During a walkdown of Fire Area 16/18, the team identified that room 805A in Fire Area 4648 lacked fire detection and fire suppression. Room 805A is the HVAC equipment room pfevIdRgwhich supplies ventilation to the Unit 2 Train A 4160V Swichgoar Room 2ETA. This ara has a modorato to high fr ladiRg ensistiRg priRnipally of ables.

The team identifiod that a rmilar onRditin also existed Fr rom 803A, which ir the HAG eqUifmoet room providing ventilaton for thnIt 1 TrI A A4160V witchgeoa Room 1 ETA in Fire Area 7switchgear room 2ETA. The team also observed that Train B cables were routed thisthrough room 805A. Many of the identified cables were in a-cable trays near the ceiling and were going from/to the cable spread room, which 16was on the same elevation-; and to/from the control room, which was above theroom 805A.

The licensee was not aware that these Train B cables passed through room 805A, and initiated Problem Investigation Process (PIP) M-03-02106 and M-03-02588. [The team identified that a similar condition also existed in room 803A (Fire Area 17), which is the HVAC equipment room supplying ventilation for the Unit 1 Train A 4160V switchgear room. The licenso had not boon awaro of all of those opposito train cables, and the' initiated PIP M 03 02106 1 ETA]. On June 10, 2003, the licensee reported that these

Charles R. Ogle - comparemcg_03_07.wpd Page 14 14 cables did not meet the separation criteria of Appendix R and represented an unanalyzed condition (Event No. 39915).

As many ar 74 oppoit a i

traw Gcab-lo aro inylhd rlated to the ractor prototon ystem. - The licensee subsequently initiated a fire watch as a compensatory measure.

Preliminary investigation by the licensee revealed that cables for primary and backup power supplies for all four reactor protection system (RPS) channels were routed in close proximity in room 805A and could be damaged during a severe fire. As many as 74 Train B RPS cables may be Involved. One consequence of this finding is that fire-induced cable damage may cause many RPS protective functions weUto spuriously go to the trip condition. SubsequentlyConsequently, a safety injection signal wouklcould be generated duo to Spurious high containment proesuro.. AThe safety injection signal we~ukcould in turn trigger a reactor trip and Phase A isolation. [At the same time, many 4npodant-main control panel instruments we-ldnecessary to achieve and maintain hot shutdown could be lost. Fr xample,, including pressurizer level and all four steam generator lvol, which are inctrumontc nococary to achieve and maintain hot shutdewni(SG) level instruments.] The licensee also stated that a-similar situatio existseffects could occur for a fire in the Unit 1 Train A switchgear room 1 ETA (Fire Area 17).

Analysis: The fa-team determined that this finding was associated with the equipment performance attribute and affected the objective of the mitigating systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events, and is therefore greater than minor. The licensee is analyzing the manner in which plant systems would be affected by fire damage to the Train B cables and is reviewing plant abnormal procedures (APs) in light of the degraded instrumentation and any automatic actions that would be initiated. Once the equipment degradations and relevant procedures are understood, the SDP will be used to determine the level of significance. When assessed in combination with the finding related to inadequate protection of AFW cables and equipment required for SSD in Fire Area 16/18 (Section.03.b.2), this finding may have potential safety significance greater than very low significance.

Enforcement: The licensee's FPP commits to 10 CFR 50, Appendix R, Section Mll.G.

Section 1ll.G.2 requires in part, that cables or equipment for one of the redundant trains of a system necessary to achieve and maintain hot shutdown (located in the same fire area) shall be ensured to be free of fire damage by one of the following: (1) separated by a 3-hour rated fire barrier; (2) separated by 20-feet or more horizontal distance with no intervening combustibles or fire hazards, and having suppression and detection; or (3) enclosure of the cables in a 1-hour rated fire barrier and having suppression and detection.

Contrary to the above, electrical cables associated with redundant trains of RPS Instrumentation necessary to achieve and maintain hot shutdown could be let due to a credible firo in one aroa as described above constitutes a violation of 10 CFR 0, Appendix R, Soction If.G.2. This section roquiroc that one train of systomc nococary to achieve and maintain hot shutdown 6hall be free of firo damage. The fact that the area procontod an exposure firG hazard to rafe shutdown

1. quipmont and did not have automatic firo detoction syctoms roprosonts a violation of 10 CFR 60, Appondix R,

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opr cL30.n ae1 I

.a 15 Section lll.F. Tho team dotorminod that this finding was associatod with the "oquipmont porFormanco attribute and affected the objoctivo of the mitigating systems cornortono to ensuro the availability, roliability and capability of ystoms that respond to initiating ovont6, and is thoroforo greater than minor. Tho finding did prosent an oporability-concorn, which the iconsoo roolvod by posting a firO watch in the aroa of concorn.

Once the liconsoo has fully analyzed tho mannor in which plant systems would have boon affocted by damage to the nopposito train" cables and roviowed the abnormal operating procoduroe in light of the dograded instrumentation and any automatic actionS that would bo initiated, the NRC will reviow this analysis. Once the equipment dogradations and rolevant procedurs are undorstood, a significanco determination procos (SDP) will be performed to dotormino the lovol of 6ignificanco. Whon assosod in combination with the finding related to inadequate protoction of cabler and equipment required for safe shutdown in Firo Aroa 16118 (also discussed in this inspection rport),

this inding may have pteRntial safoty 6igRificGanco gatrao' low sigRificanRc.

Enforcomont: As doscribhd above, the fidiRng is a violatioR of ApprndiX R requirements of greator than minor significancodamaged during a fire in room 805A (Fire Area 16/18).

Pending determination of the systems affected and the safety significance, the finding is Identified as URI 50-369, 370/03-07-02, Failuro to ProtOct Roactor Frotoction System Cables Rerults in Loss of Roquired ShutdownInadequate Separation and Protection of Cables Associated With Redundant Trains of Instrumentation Located in the Same Fire Area.

2.

Inadequate Protection of AFW Cables and Equipment-and-Cables Required for Safe Shutdown

==

Introduction:==

A finding was dentified in that physical protection of the assiate4 electrical cables feFassociated with valve 2CA0042B (2B motor driven AFW pump discharge supply to steam gRaterSG 2D) did not meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Instead, the licensee substituted tho "se-Gfd a local manual operator action, which had not received prior NRC approval, to achieve and maintain SSD. This is a URI pending completion of the SDP.

DescriDtion: On April 2, 2003, t The licensee identified (April 2003) that MNS relied on local, manual operator actions outside the MCR for SSD in nonRaltornativo;Gion-dedicated shutdown fire areas (i.e., areas designated as complying with 10 CFR 50, Appendix R, Section II.G.2) and the. These local, manual operator actions did not have prior NRC approval. The licensee documented this issue in PIP M 03 02314 M-03-0231 1. The team reviewed the local, manual operator actions for the Appendix R,Section III.G.2 fire area selected for this inspection (Fire Area 16/18).

The team found that the associated electrical cables for Train B valve 2CA0042B were located In the Unit 2 Train A 2ETNEelectrical Ppenetration Rroom (Fire Area 16/18) without adequate spatial separation or fire barriers. Tho licenseo's SSA stated that do nergizing this valve aftor vorifying that it was open was a time critical action because spurious closure of this valve wold limit the socondary heat sink to only one steam genorator instead of tho two required for SSD. Howovor, Rather than providing-adequate physical protection for redundant trains of equipment/systems necessary to

Charles R. 0gle - cornare_mcg03_07.wpd Page 16 16 achieve and maintain SSD (as specified for Appendix R,Section II.G.2 areas), the licensee substituted the use of a manual operator actions outside the MCR. The licensee's SSA stated that de-energizing this valve, after verifying that it was open, was a time critical action because spurious closure of this valve would limit the secondary heat sink to only one SG (rather than the two required to achieve and maintain SSD).

The use of local manual operator actions, in fire areas designated as complying with the provisions of Appendix R,Section III.G.2, requires prior NRC review and approval.

T:heseThis local, manual operator actions had not received NRC approval.

Analysis: The team determined that this finding was associated with the "equipmentequipment performance! attribute of the mitigating systems cornerstone. It affected this cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events, and is therefore greater than minor. When assessed in combination with the inadequate reactor prtectiRo ystemRPS cable separation finding (also discussod in thiS inspoction roportSection.03.b.1), this finding may have potential safety significance greater than very low significance.

Enforcement: The licensee's Firo PFrotoVtion PregramFPP commits to 10 CFR 50, Appendix R, Section III.G.Section III.G.2 statesrequires in part, that-,

"w...Whoe cables or equipmentiisluding associatod non safEty circuits that could pro'ent onoMtion or Gause malonOration due to hot shnrt oon circuits, or shortS t ground, ef for one of the redundant trains of a systems necessary to achieve and maintain hot shutdown Gnditions are(located withinin the same fire areaGutside) shall be ensured to be free of primary centainmont,fire damage by one of the following meaRs of ensuring that ono of the redundant trains is froe of firo damage shall bo provided: (1) separation of cables and equipment of redundant trains by a firo barrier having a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rating; (2) separation of cables and equipmont of redundant trains by a: (1) separated by a 3-hour rated fire barrier; (2) separated by 20-feet or more horizontal distance ef re-than 2feet with no intervening combustibles or fire hazards. In addition, firo detectors and an automatic fire upprors6ionv system shall bo installod in the fire area;, and having suppression and detection; or (3) enclosure of Gablos and equipment of ono redundant train in afire barrierhaving a 1 hou--r rating. lIaddition, firod dotoctorc nd an automati fire suppression ystem shall be installed in the fire area. the cables in a 1-hour rated fire barrier and having suppression and detection.

Contrary to the above, on May 23, 2003, the team found that the licensee failed to protect electrical cables efassociated with redundant equipment located within the Unit 2 Train A Switchgear Room/tElectrical Pelectrical penetration Rroom (Fire Area 16/18) with an adequate barrier or to provide 20 feet of separation. Instead, the licensee used a local manual operator action, which had not received prior NRC approval, to achieve and maintain SSD. Pending determination of the finding's safety significance, this finding Is identified as URI 50-370/03-07-053, FailurE to Provide AdoquatoUse of a Local Manual Operator Action in Ueu of Providing Physical Protection for Cables of Redundant Safe Shutdown Equipment in Fire Area 16/18.

.04 Alternative Post-Fire Safe Shutdown Capability

a.

Inspection Scope

I Charles R. Ogle - comparemcg.03_07.wpd Page 17 17 The team reviewed the licensee's procedures for fire response, abnFFral PFGeeduosAPs for altomativo shutdown (ASD)DSD, and the licensee's Appendix R manual action roquiromontc analysoefire area failure analysis and compliance strategy for a fire in the seleeted Fire Areas 4, 13, and 24. The team also walked down selected portions of the procedures in the plant. The reviews focused on ensuring that the required functions for post-fire safe shutdown and the corresponding equipment necessary to perform those functions were included in the procedures. The review also included assessing whether hot and cold shutdown from outside the MCR could be implemented, and that transfer of control from the MCR to the standby shutdown facility (SSF) could be accomplished within the performance goals stated in 10 CFR 50, Appendix R, Section III.L. The components listed in Section 4R05.03.a. of this inspection FeportIR were also reviewed in relation to altrnative potfiro safo ehutdowADSD capability. The team reviewed the most recently completed surveillances for selected instruments required during SSS operation to verify that these surveillances were being completed in accordance with MNS SLC 16.9.7, Standby Shutdown System.

The team walked downs focused on ensuring that the DSD procedures to determine if they could easoRably be performed within the required times-given the minimum required staffing level of operators-aA, with or without offsite power available. The team also reviewed the electrical isolation of selected motor operated valves from the control room to verify that operation of the SSS from the SSF, and other remote plant locations, would not be prevented by a fire-induced circuit fault. Tho objoctiVo of those roviov.'c *wr to ro that tho poct fire 6cat9 ohtdWAn analical approach, 6afo shutdon equipment, and proodrocs woro Gonsistont and Gompliod Wth tho Appondix D roator pr rmanoe ritria fr safe hutdown.

b.

Eindings

1.

Requirements Relative to the Number of SDurious ODerations that Must be Postulated

==

Introduction:==

An unreolved item wasThe team identified an issue involving the number of concurrent spurious operations associated with a particular component or set of components that must be postulated RlktieR during SSD analysis of tho unrooed 4.-isa fire area. This issue is a URI pending review byof NRC &tafguidance in this area.

==

Description:==

The licensee's fire protoction analysirSSA included the concept that only one spurious operation due to fire damage need be postulated. This concept became evident during review of the pressurizer PORVs. There are three sets of PORWPQRVPORVs and PORV isolation valves on the pressurizer of each unit. Should operators in the control room become aware of a fire in any aeaef-the plant twugharea (from a fire alarm or the plant communications system), they would respond by following the instructions in abnormal pimplementing Procedure AP/OA/5500/045, Plant Fire. Depending on the fire location, pProcedure AP/0/A5500/045 directed the operator to close the PORV isolation valves within ten minutes. The basis for this time critical action is the licensee's assumption that spurious opening of the PORV, or damage to the isolation valve circuit would not occur in the first ten minutes of a fire being detected. Then wWith the WGeeIkisolation valve closed, it would then take two spurious operations to breach the RCS pressure boundary (i.e., namely on bkthe isolation valve opening and its associated PORV also opening).- Theis concept of

i Charles R. Oale - COMDare mca 03 07.wDd Paae 1 8 9k Chre R Q-= - -- m-re inc- 03 7wo

-ae 18---

i--

18 postulating only one spurious operation noed be postuated-meant that closing the bleekisolation valve was sufficient JRitselfto ensure the desired rosultRCS pressure boundary integrity. The licensee considered that there was no need to take any other action such as de-energizing the isolation valves after they vwerit was closed.

FApplication of this concept wasis not noeessaily consistent with NTNRC's cable protection requirements for protection of cablosof Appendix R,Section III.G.

The team reviewed the control circuits and cable routing information for valve 2NC34A, pressurizer PORV 2NC34A, and-2NC33A its associated isolation valve 2NC33A. They observed that cables for both the PORV and isolation valve arewere routed ithrough Fire Areas 13, 16/18 and 24. Whon the control circuit for the PORV is analyzed and coRid0orin that tho Galoles aoe arord typo Gablos (oXpt ir tho ontrol room) oro Gan seGnclGudeThe team determined that, for these three fire areas, spurious opening of the PORV could only occur for the firo in Firo Aroa 21, the control room. Concidring this information, the tam postulated tho following cOnario. A firo start in the control room. Oporator closeo tho isolation valves por procedure APO/ON5600!016 within ton minutoc. Lator, isolation valve 2NC33A spuriously opens duo to a firo induced short circuit. Operators tako no action to counter tho spurious oponing of tho isolation valve because they have no information that it occurrod. Subsoquently PORV 2NC34A spuriously opens duo to a fire induced short circuit. At this point, it would be possibl to CGoco the PORV by opOning the appropriato circuit breaker at tho 126 VDC distribution panel. This would take timo, and it is not covored by tho firo rosponco procoduro.

Boforo tho PORV can be ro closed, the fire has progressod and the docision is mado to abandon the control room and shutdown using tho SSS. The PORV would now bo Glosed by operating the control roomSSS transfer switch as directed by abnormal procedure APAW55\\N400/021, Loss of Plant Control Due to Firo or Sabotago. The situation now is that the PORVfisolation valves were oponed for a period of time and the RCS is may not be at normal lvol and proscuro. Tho standby makeup pump has rolatively low capacity and may not have the capacity to maintain hot hutdown in this seenario, and RCS variable paramotors may be outside tho roquiromonts of Appondix R, i.o. outside the range predicted for a 10c6 Of off6ito powor. For oxamplo, an open PORV following a roactor trip could result in pressurizer leve! lower than that prodictod frF trip rausod by a loss Of oft poWor.

bluk:: Trh tam was rnt cortain wheothr th lonRseo's analysis Of circuits for spurious E Nopration was consRitont th tho ruromorants for indeopndoncn n Abloc, yctroms or crmponRts in theon undor onsidoration as stipulated by Appendix R, lll3 nd 11.

In tho example of tho PR\\'Rs deosribod ab--

If MCR fire (Fire Area 24). If more than one spurious operation weGdwere to occur, the dedicated shutdown capability (i.e., the SSS) would not be independent from the GGRtFGI FGGMMCR in that, during a fire in the control FeemMCR, pressurizer level may not remain within the indicating range which could result in conditions outside efthose specified in U4LAppendix R,Section III.L.

Analysis: The team determined that this finding was associated with the equipment performance attribute of the mitigating systems cornerstone. Because it affected this cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events, this finding is greater than minor. If more than one spurious operation must be considered then there would be a violation of Appondix R

Charles R. Ogle - comparemcg 03_07.

Page 19 19 roquiromonte having moro than minor 6ignificanco. The quipnont roliability objoctivo of the cornerstone of mitigating systems and barrior integrity could bo affected.

were to occur, the dedicated shutdown capability (i.e., the SSS) would not be independent from the MCR in that a fire in the MCR could result in pressurizer level not remaining within the indicating range.

Enforcement: In the case of the PORV/irolationPORV and PORV isolation valve circuits, operation of the SSS may not be independent of the fire area as required by Appendix R,Section III.G.3 depeonding on whothor moro than ono purious oporation must be postulated. Review of this matter by the NRC will determine whether a violation has occurred. If a violation has occurred, the significance will be dotorminod. T Pending review of NRC guidance in this area, the issue is identified as URI 60-36950-369/03-07-03, 370/03-07-034, Requirements Relative to the Number of Spurious Operations tThat mMust be Postulated.

2

'.'alvo 2CA0007A 2. Auxiliary Feedwater Valve 2CA0007A not Included in Safe Shutdown Analysis

==

Introduction:==

A finding efwith potentially greater than very low safety significance was identified in that a valve in the auxiliary foodwator ryctomAFW suction supply valve 2CA0007A, which could spuriously operate during a MCR fire, was not included in the cafe hutdown analysis and it could puriously dloGe duo ta firo in the main control foeR';SSA. Spurious closure of this valve could damage the turbine driven auxiliary feedwater (TDAFW) pump, thus seriously degrading the GeF9-eseikalsecondary decay heat removal function of the safe hutdown systomSSS. This is a URI pending completion of the SDP.

==

Description:==

Valve 2CA0007A is a motor operated valve In the suction flow path from the 300,000 gallon auxiliary foodwatorAFW storage tank to the turbino driven auxiliary feedwatefTDAFW pump. The valve is open during normal plant operation. Valve 2CA0007A is important to safe shutdown for fire areas where the safo hutdown cystom (SSS) will be used. The importanGo is dorivod froG fact that because the SSS usesrelies on the TDAFW pump for secondary decay heat removal and potential for Spurious closure of the valve. The team found that the safe shutdown analysis for Unit 2 did not recognize valve 2CAO0O7A. It was not listed in Appondix E, list of important oquipmont, nor Appondix F, list of potential problem cabloc.

One rccnario could bo a fire starts in the control room which loads to a plant trip and lorc of fftio poWer. I this taco, the T ADAP n

pump would rciyo ar autoematic strt from the LOHI on afety raRta hai' logic Or poctibly low i

eaams 9nerator leol do to lA Af th fedwator pump. Even though the salo shutdow aRalytir Fr a fire R the control romw ultimately rlioc on the SSS, op9ratGrF ma romkin in the rontrol room if thoy blievo the plant i6 still undor control. Tho TDAF' pump could bo running and taking Ucto fm tho auxiliary fedwator ttrago tank with flow through CO 7A.

Sineo onrol Wiser to the pen/deose caotrol switch fo thit absE rU ir the ontrol oom (i;

iRgle coRductor plug able, bundled i group of approimatoly 3 Wiro), the valve could 6puriously coEo duo to fire induced short circuit between two of the wirot.

Spurious closure of the valve would immediately reduce suction pressure and quickly shut off all normal AFW flow through the pump. A66uming that the TDAFA' pump It

C Charles R. Ogle - comparemcg_03_07.wpd Page 20 20 damaged by spurious closuro of 2CA007A and if plant conditions dotorioratod duo to progressing firo in the control room forcingClosure of this valve could cause severe damage to the pump if automatic transfer to the alternate suction sources does not initiate within sufficient time. For e severe fire in the MCR requiring evacuation and transfer of plant shutdown to the SSS, the ability to remove decay heat would be seriously degraded.

Bosidos the control room, thoro aro opon/closo switchos for this valve at auxiliary-foodwator panol 2A and tho auxiliary foodwator turbino control panel (2AFPT). Cablo 2*GA517 runs botwoon area trminal cabinet 2ATC2 and the auxiliary foodwator panol 2A, and it runs through firo area FA 4. Cable 2*CA51 a runs botwoon aroa trminal-cabinot 2ATC2 and panel 2AFPT, and it runs through firoeroa FA 4. ablo 2*A557 contains powor and control for the valve, and reprosents a potential for spurious operation of the valve. Thoroforo a fire in FA 1 could also rsult in surious closuro of if the TDAFW pump were damaged. The team found that the SSA did not include valve 2CA0007A. This could load to problems similar to that doscribed abovo for tho control room firo. It is not oxpoctod that a fire in FA 4 would load to a loss of offsito power.

Howovor, a problom cenario could bo as fllows: If the fire becomes sovoro and the docision Is made to useo the SSS, proceduros diroct tho oporator to trip the normal foodwator pump. This could cause low steam generator lvl which in turn willaut-start the TDAFW pump. If 2CA0007A has already spuriously closed, the pump has no through flow upon startingThe valve was not listed in Appendix E, Unit 1 and Unit 2 Safe Shutdown Equipment; nor Appendix F, Fire Area Failure Analysis and Compliance Strategy, of the SSA (MCS-1465.00-00-0022, Design Basis Specification for Appendix R).

The licensee initiated a correctivo action documontPlPs 02084, M-03-02118, and M-03-02311 for this issue, PIP M 03 02081, nd4hey took prompt action to este-operability. Thoy rvised AP 24prevent spurious operation of this valve. Procedure AP/O/A/5500/045 was revised to specify that the operator chock that valve 2CA007A is open and removo power from 2CAO07Aonsure, within the first ten minutes of a4iFean active fire, that valve 2CA0007A was open and then remove power from 2CA0007A.

The team noted that system design provided for automatic transfer to alternate suction sources initiated by pressure switches In the TDAFW pump suction line. There were three separate alternate suction flow paths. Path 1 was through valves 2CA1 61 C, 2CA1 62C and ORN4AC; Path 2 was through valves 2CA086A and 2RN069A; and Path 3 was through valves 2CA1 16B and 2RN1 62B. However, key information related to these automatic transfers was not available to the team during the inspection.

Information was subsequently provided to the team, however, this information has not yet been fully reviewed.

Analysis: The team determined that this finding was associated with the "equipmentequipment performance! attribute and affected the objective of the mitigating systems cornerstone to ensure the availability, reliability and capability of systems that respond to initiating events, and is therefore greater than minor. For a severe fire in the MFGG FComMCR, the contro rOOMMCR would be abandoned vacuated and the safe-shutdown aGiitySSF would be used to achieve and maintain hot shutdown. The Gafe-shutdoWn facility Folios on the turbine drivon auxiliary foodwator pump for the decay heat

Charles R. Ogle - compare-mcg_03_07.wpd Page 21 21 removal function. Wih the decay heat romoval function oriously dogradod and othor mitigating systems potentially affocted by a sovoro control room firo or Firo Aroa 4, the finding was also determined to had-ave potential safety significance greater than very low. The team was awaro that system dosign provided for automatic transfers to altomato suction sources initiated by prossuro switchos in the pump suction line. Thoro woro three separate alternate suction flow paths. Path 1 was through valves 2CA1 61 C, 2CA1 62C and ORNIAC; Path 2 was through Valvos 2CA086A and 2RN06OA; and Path 3 was through valvos 2CA1 1 6 -and 2RN1 62B. Howovor, kRoy infoRmation rolatod to thoso automatic transfors was not available to the team at tho tim of this inspoction-roport issuance. One quostion was whether the automatic transfor on low suction prossure would occur fact enough to protoct tho pump for tho case significance because the SSF relies on the TDAFW pump for decay heat removal, and the decay heat removal function would be seriously degraded if the TDAFW pump were damaged due to closure of valve 2CAOOO7A Glosing since this valve was oso to tho pump. In answOring this guoction, tho liconcop statod, and prosontod same information, that a few ovonts had occurrod ovor the years whore suction valvos wore inadvortontly Glosed While mnotor drive'n ARN pumAps Wero unnig an h ump was not damaged. Dtails of these events and similarity of the motor driven and turbine drivOn pumps havo not boon roviowed by the team. Secondly, the liconso provided information to the tam, subsequent to the inspocion, on the routing of all the valvos involved in the automatic transfers. Howovor, thic information has not yAt boon fully roviowod by tam to dotormine whothor or not tho transfors could be affected by the amo firo which caused tho 2GAO7A valvo to spuriously lo1.

This informatlon would ho RRdod to onmaplot tho 6igRifiaRno detorminatioR processr,.

Enforcement: 10 CFR 50.48 states, in part, that each operating nuclear power plant must have a fire protection program that satisfies Criterion 3 of 10 CFR 50, Appendix RT-Sotion.B. rquiros that aA. MNS Unit 2 Operating License NPF-17, Condition 2.C.(7) states, in part, that the licensee shall implement and maintain in effect all provisions of the approved FPP as described in the UFSAR for the facility, and as approved In the SER dated March 1978 and SER Supplements 2, 5, and 6 dated March 1979, April 1981, and February 1983, respectively, and the safety evaluation dated May 15, 1989.

The McGuire FPP, which includes the SSA (MCS-1465.00-00-0022), states in part, that the FPP implemented the philosophy of defense-in-depth protection against fire hazards analysis shall be porformod by qualified fire protoction and reactor systoms onginoors to determine the onsequRnro f fir in aRy lcation f tho plant on the ability to safely shutdown the roactorand effects of fire on SSD equipment. It further states that the SSA performed for MNS considered potential fire hazards and their possible effects on SSD capability. The licensee's anaysieSSA designated the MCR (Fire Area 24) and Fire Area 4 as dodicatodealtomativodedicated shutdown areas. Appendix R, Section 1Il.G.3 requires that the dodicatod/altornativsalternative/dedicated shutdown capability, and its associated circuits, be independent of cables, systems or components in the area under consideration.

Contrary to these requirements, valve 2CA0007A was not included in the fihaiards-apalysiGSSA resulting in the altornativo/dodicatoddedicated shutdown system (SSS) not being independent from Fire Area 4 adArea 24, in that, a fire in these areas could

Charles R. Ogle - comparemcg_03_07.wpd Page 22 22 result in spurious closure of theis valvo. This in tr could load to damage to the turbine drio auxiliary foodwator pump Which waG required fr alternativo shutdon using the SSS and damage to the TDAFW pump. Pending determination of the safety significance, this finding is identified as URI 50-370/03-07-085, Spurious Closure of Valve 2CA0007A Could Lead to Damage of the TDAFW Pump.-

.05 Operational Implementation of Post-Fire Safe Shutdown Canabilitv

a.

Inspection Scone The team reviewed the operational implementation of the altomativo shutdownSSD capability for a fire in Fire Areas 4, 13, 16/18, or 24 to verify that: (1) the training program for licensed personnel included alternative or dedicated safe shutdown capability; (2) personnel required to achieve and maintain the plant in hot standby following a fire using the SSS -could be provided from normal onsite staff, exclusive of the fire brigade; (3) the licensee had incorporated the operability of atefnativededicated shutdown transfer and control functions Into plant TS and/or SLCs; and (4) the licensee periodically performed operability testing of the alteffativededicated shutdown instrumentation, and transfer and control functions. The team reviewed abROFma-pProcedures AP/1/A/5500/24 and AP/2/A/5500/024, Loss of Plant Control Due to Fire or Sabotage, and AP/0/A/5500/045, Plant Fire. The reviews focused on ensuring that all required functions for post-fire safe shutdown, and the corresponding equipment necessary to perform those functions, were included in the procedures.

Tho objective of this reviow was to assure that the safe shutdown equipment, 6hutdown procedures, and the post firo safe shutdown analAical approach wAre consistOnt and satisfied tho Appendix R reactor performance criteria for safo shutdown.

b.

Findings The licensee identified that local, manual operator actions outside the MCR were used in lieu of physical protection of equipment and cables relied Goupon for SSD during a fireT without obtaining prior NRC approval. FA specific findings related to this issue afefor Fire Area 16/18 is discussed in Section R0&03.b.2 of this inspection roport for Firo Arpa 16!481R.

The team identified a URI regarding the adequacy of the licensee's method for controlling RCS pressure during operation from the SSF in the event of a fire. -

During review of procedures AP/1/A/5500/024 and AP121A/5500/024, the team questioned the adequacy of the 70 kilowatts (kwW) pressurizer heater capacity (per unit) powered from the SSF to maintain and control RCS pressure in hot standby during a fire in plant areas which require use'of the SSS. The question WaS aised whon the team obsered that a A procedural note in both AP/1/A/55001024 and AP/2/A/55001024 provided guidance to the operators which stated that it was acceptable to allow the RGSpressurizer to go water solid in order to maintain subcooling, and7 with the RGSpressurizer water solid, the reactor vessel head vents would be used to control pressure. Tho team quostionod why this guidance was in those procedures. Allowing the pressurizer to go water solid for controlling RCS pressure during hot standby conditions while operating from the SSF was not consistent with Appendix R, Section lll.L, for altoefativededicated shutdown capability, nor the design basis description for the SSF as stated in the licensee's letter to the NRC dated March 31, 1980. Also, solid

I Charles R. Oale - comDaremcQ_03_07.wvd Page 23 Chale R. Ocl oprmcj37wdce2 23 plant operation from the SSF for controlling RCS pressure was neither reviewed nor discussed in any NRC SER/SER Supplements relative to acceptability of the SSF design for altefativededicated shutdown capability. The team requested information from the licensee (e.g., analyses, calculations, etc.) which demonstrated the following:

Adequacy of the 70 kwW pressurizer heater capacity powered from the SSF for maintaining and controlling RCS pressure in hot standby.

AfeValidity of the assumptions for pressurizer heat loss stated in the October 21, 1980, letter still valid (based on insulation degradation and/or degraded capacity of the heaters -powered from SSF) for assumig current pressurizer heat loss and for determining when the heaters will be needed.

SMP capacity to achieve and control solid plant operation from the SSF within the required time to maintain subcooling.

Operator training (Psjob performance measures, simulator, etc.) on solid plant operation from the SSF.

The licensee indicated that there were no specific calculations documented which provided the basis for the number of heaters to be powered from the SSF. The licensee further stated that there was no calculation which demonstrated the performance capability of the SMP during solid plant operation from the SSF. The licensee also indicated that training provided to operators on solid plant operation from the SSF consisted primarily of classroom discussions and tabletop wa*k-thFuhsdiscussions of pProcedures AP/1/AI5500/024 and AP/2/A/55001024. The team concluded that sufficient information was not provided to resolve the questions raised above nor to determine the licensee's ability to safely operate the SSF with the pressurizer in a water solid condition during fire events in areas where the SSF is used to achieve SSD.

FPending further NRC review of additional licensee information, this issue is identified as URI 60 36950-369/03-07-04, 370/03-07-046, Methods for Reactor Coolant System Pressure Control During SSF Operation, pending further NRC roviow of additional liGono n frmation.

.06 Communications

a.

Inspection Scope The team reviewed plant communication capabilities to verify that they were adequate to support unit shutdown and fire brigade duties. This included verifying that site paging (PA),portable radios, and sound-powered phone systems were consistent with the licensing basis and would be available during fire response activities. The team reviewed the licensee's communications features to assess whether they were properly evaluated in the licensee's SSA (protected from exposure fire damage) and properly integrated Into the post-fire SSD procedures. The team also walked down sections of the post-fire SSD procedures to verify that adequate communications equipment would be available to support the SSD process.

b.

Findings

a Charles R. Ogle - compare-mcgO3O07.wpd Page 24 24 No findings of significance were identified.

- i Charles R. Ocile - comDaremcci_03_07.wPd Pacie 2' 0

Chre R_ Ogl-coprecOO d

_ae2 25

.07 Emeraencv Lighting

a.

Inspection Scope The team compared the installation of the licensee's emergency lighting systems to the requirements of 10 CFR 50, Appendix R, Section lll.J, to verify that 8-hour emergency lighting coverage was provided In areas where manual local operator actions were required during post-fire SSD operations, including the access and egress routes. The team's review also Included verifying that emergency lighting requirements were evaluated in the licensee's SSA and properly integrated into the post-fire SSD procedures. During pSteam walk downs of the selected areas where local, manual operator actions would be performed, the team incpoctod area emergency lighting units (ELV)were inspected for operability and cheked-the aiming of lamp heads was checked to determine if adequate illumination waswould be available to correctly and safely perform the actions directed by the procedures.

b.

Findinas No findings of significance were identified.

.08 Cold Shutdown Repairs

a.

Inspection Scope The team reviewed the licensee's SSA and existing plant procedures to determine if any repairs were necessary to achieve cold shutdown, and if needed, the equipment and procedures required to implement those repairs were available onsite.

b.

Findinas No findings of significance were identified.

.09 Fire Barriers and Fire Area/Zone/Room Penetration Seals

a.

Inspection Scope The team reviewed the selected fire areas to evaluate the adequacy of the fire resistance of fire area barrier enclosure walls, ceilings, floors, fire barrier mechanical and electrical penetration seals, fire doors, and fire dampers. This was accomplished by observing the material condition and configuration of the installed fire barrier features, as well as, construction details and supporting fire endurance tests for the installed fire barrier features, to verify the as-built configurations were qualified by appropriate fire endurance tests. The team also reviewed the fire hazards analysis to verify the fire loading used by the licensee to determine the fire resistive rating of the fire barrier enclosures. The team also reviewed the design specification for mechanical and electrical penetrations-,, fire flood and pressure seals, penetration seal database and Generic Letter (GL) 86-10 evaluations and the calculation for the technical basis of fire barrier penetration seals to verify that the fire barrier installations met licensing basis commitments.

i Charles R. Oale - COMDare mca 03 07.wDd Paae 26 Charles R. QAle -I co-oarew e 26 26 The team reviewed fire barriers shown on the fire plan drawings for the selected fire areas. The statleteam noted that MNS has eliminated selected fire barriers from the approved fire protection program and designatesd these fire barriers as "Sealed Firewall

- Non Gemmitted". Committed." These barriers are no longer included in any surveillance and testing program. Therefore, doors, dampers, fire proofing, etc. that exist in these declassified barriers are no longer included in any station surveillance procedures and effectively cannot be relied upon for the fire protection program. Two walls associated with Fire Area 4816/18 have been declassified. The wall between the SUnit 2 switchgear Rroom 2ETA (Fire Area 18) and the EUnit 2 electrical Ppenetration Afearoom (Fire Area 16) was declassified in Revision 9 (2000) ad4. The wall between the SUnit 2 switchgear Rroom 2ETA (Fire Area 18) and the Unit 2 HVAC Eequipment Afearoom 805A (Fire Area 18) was declassified in ReyienRev. 3 (1982). The team roquostod the Liconspe to provide the onginoring analyse that supports the dhclasciicationof those barriFr.

For the purposes of the inspection of Fire Area 18, the Selectrical Penetration Arearoom (Fire Area 16) was included in the inspection plan because the fire wall separating these areas has been declassified and is no longer a "Fire Sealed - NRC Committed" fire barrier. The similar wall at Unit 1 Room 803A was also declassified from a "Sealed Firewall - NRC Committed" to a "Sealed Firewall - Non Committed."

The team walked down the selected fire zones/areas to evaluate the adequacy of the fire resistance of barrier enclosure walls, ceilings, floors, and cable protection. The team selected several fire barrier features for detailed evaluation and inspection to verify proper installation and qualification. These features included fire barrier penetration fire stop seals, fire doors, fire dampers, and fire barrier partitions, and Thormo Lag e9octrical raceway fiFe barrier system (RFBS) eclosuere.

The team observed the material condition and configuration of the selected fire barrier features and also reviewed construction details and supporting fire endurance tests for the installed fire barrier features. This review was performed to verify that the observed fire barrier penetration seal aidERFBS-onfigurations conformed with the design drawings and tested configurations. The team also compared the penetration sealand ERFBS ratings with the ratings of the barriers in which they were installed.

The team reviewed licensing documentation, engineering evaluations of G1rhie-LettefGL 86-10 fire barrier features, and NFPA code deviations to verify that the fire barrier installations met design requirements and license commitments. In addition, the team reviewed surveillance and maintenance procedures for selected fire barrier features to verify the fire barriers were being adequately maintained.

b.

Findings No findings of significance were identified.

.10 Fire Protection Systems. Features, and Equipment

a.

Inspection Scope

i Charles R. Oale - compare-mcq-03-07.wpd Page 27 CarlsRQe-cmaemcOO.p ae2 27 The team reviewed UFSAR Section 9.5.1, Design Basis Spocification for Firo Protoction, Firo Protoction Code Deiations, and Administrativothe fire protection design basis specification, fire protection code deviations, and administrative procedures used to prevent fires and control combustible hazards and ignition sources. This review was performed to verify that the objectives established by the NRC-approved FPP were satisfied. The team also toured the selected plant fire areas to observe the licensee's implementation of these procedures.

The team reviewed the adequacy of the design and installation of the automatic wet pipe sprinkler system protecting the RN pumps in Fire Area 4. Team members performed a walk down of the system to ensure proper placement and spacing of the sprinkler heads and the extent of the sprinkler head obstructions. Selected engineering evaluations for NFPA code deviations were reviewed and compared aga4ixewith the physical configuration of the system. The team reviewed the sprinkler system hydraulic calculations for this system to ensure that the system could be supplied sufficient pressure and volume utilizing the two by-pass lines without opening the deluge valves.

The team also inspected one of the by-pass lines located in an outside pit to determine the piping and fitting equivalent length to confirm the accuratonosaccuracy of the design input to the RN pump calculation. The team reviewed the fire protection code deviations calculation for automatic suppression systems relative to the selected fire areas.

The team reviewed the adequacy of the design and installation of the automatic detection and alarm system for the selected fire areas. This was accomplished by reviewing the ceiling reinforcing plans and beam schedule drawings to determine the location of ceiling bays. After the ceiling bay locations were Identified, the team conducted a plant tour to confirm that each bay was protected by a fire detector in accordance with the Code of Record requirements -

NFPA 72E, 1974. Field tours were conducted in fire areas 13, 16/18 to confirm detector locations. Minor modification package MM-1 2907 was reviewed where 10 new detectors were added to Fire Area 13 to conform the detection system to NFPA 72E location requirements.

The team reviewed the fire protection code deviations calculation for automatic detection systems relative to the selected areas to determine if there were any code deviations cited for the selected fire areas.

The team reviewed the fire protection pre-plans and fire strategies to ensure that hose locations could sufficiently reach the selected fire areas for manual fire fighting efforts.

Hose stations in the selected area were inspected to ensure that hose lengths depicted on the engineering documents were also the hose lengths located in the field. This was done to ensure that manual fire fighting efforts could be accomplished in the selected fire areas.

b.

Findings No findings of significance were Identified.

4.

Other ActivitlocOTHER ACTIVMES

Charles R. Ogle - compare-mcg03_07.wpd Page 28 28 4OA2 Problem Identification and Resolution

a.

Inspection Scope The team reviewed a sample of licensee audits, self-assessments, and PIPs to verify that items related to fire protection and to SSD were appropriately entered into the licensee's GAPcorrective action program in accordance with the MNS quality assurance program and procedural requirements. The items selected were reviewed for classification, appropriateness, and timeliness of the corrective actions taken, or initiated, to resolve the issues. Included in this review were PIPs G-99-00110, M-99-01884, M-99-01886, M-03-01675, and minor modification MM-12907 related to the McGuire Fire Protection Functional Audit SA-99-04(MC)(RA)(FPFA).

In addition, the team reviewed the licensee's applicability evaluations and corrective actions for selected industry experience issues related to fire protection. The operating experience4(QE) reports were reviewed to verify that the licensee's review and actions were appropriate.

b.

Findings One liconseo identified finding (rolatod to the use of manual oporator actions in Firo-Area 16M 8 without prior NIC approval) involved a violation of NRC roquiromonts. The onforcomont onsidortions for this violation are discussod in Soction 1 R05.03.b.2 of this inspection rOport.

Th tam obsorvod that the adequacy and timolinoss of corractive actions to address the findings from the Firo PFrotction Functional Audit SA 99 4(MC)(RA)(FPFA) rogarding fire detoction in the Battory Rooms (Firo Aroa 13) wero not commonsurate with the risk significanc aociatod with a fire in this area. The liconseo's IPEEE identified that a fire in the Battory Rooms ranked as the top contributor to CDF. The fir detertioR findingsNo findings of significance were Identified in a 1999 lioonoo self initiated tchnical audit (SITA) SA 99 01. Howovor, the nitial minor modification (MM 1207) scOpo was inadequate in that only two additional detactors Woro to be installed in the battery rooms (instead of nino roquired to comply with the NFPA Codo).

Additionally, tho modification implementation dato was postponed at loast twio. Also, tho lisoncoc had initiated PIP M 03 01676 (dated April 10, 2003) rgarding detectors not being installed in accordanco With NFPA codes. Whon the battery rooms fire area wore olaod bythe toam duiringto pho P-inpction Wformation gathorin vit, the toam-noted that the moedification was roveiod to install th rquirod numb:or of dtoctors and rocoiyod high prority status r implementation. Tho attery Room dttrs WOro inAltalled p;o to the fr st weok of the onsito nspoction (May 5 0, 203).

40A5 Other Activities

.01 (Closed) URI 50-369370/00-09-04: Adequacy of the Fire Rating of Mineral Insulated Cables in Lieu of Thermo-Lag Electrical Raceway Fire Barrier Systems The NRC had opened this URI for further NRC review of the adequacy of the fire resistance rating of certain mineral insulated cables that the licensee had installed. The licensee had replaced an inadequate 3-hour Thermo-Lag fire barrier with mineral insulated cables7 for charging pump 1A-in the Unit 1 tTrain B switchgear room.

, Charles R. Ogle - comparemcg_03_07.wpd Page 29 29 However, the adequacy of the testing of the mineral insulated cables, to assure their 3-hour fire resistance ability, had not been reviewed by the NRC.

The inspectors reviewed the NRC Safoty Evaluation Roport (SER) of January 13, 2003, on the licensee's use of mineral insulated cables and also reviewed the licensee's 10 CFR 50.59 safety evaluation for the modification. The NRC SER evaluated the licensee's installation and fire testing of the mineral insulated cables and concluded that the licensee had adequately demonstrated that the protection provided by the mineral Insulated cables in the specific application was equivalent to the protection provided by a 3-hour rated fire barrier. The NRC SER further concluded that this change to the approved fire protection program did not adversely affect the ability to achieve and maintain safe shutdown In the event of a fire and, therefore, did not require prior approval of the NRC. The Inspectors concluded that the licensee's 50.59 safety evaluation for the change had adequately considered that the change did not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Consequently, the licensee's installation of mineral insulated cables was not a violation of NRC requirements. This URI Is closed.

40A6 Meetings On May 23, 2003, tThe team presented the interim inspection results to yeuMr. D. Jamil and other members of yeuF staftfwhethe licensee's staff on May 22, 2003. A final exit meeting was held via telephone with Mr. G. Peterson, and other members of the licensee's staff on July 2, 2003, to present the final results of the inspection. The licensee acknowledged the findings presented. Th tam confrmod that pProprietary information is not included in thithe inspection report.

- I Charles R. Oqle - compare-mcq-03-07.wi)d Pane 30 b_ Cre R ~

- copremaO

.wp Pa 30 30 SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee Personnel D. Bailey, Mechanical and Civil Engineering (MCE) - Civil J. Boyle, Training Manager S. Bradshaw, Superintendent of Operations H. Brandes, Consulting Engineer, General Office Fire Protection Program J. Bryant, Regulatory Compliance Engineer M. Dicks, Engineer, Reactor and Electrical Systems (RES)

B. Dolan, Safety Assurance Manager J. Hackney, Operations T. Harrell, McGuire Station Manager D. Henneke, Engineer, General Office Probabilistic and Risk Assessment Group D. Herrick, Civil Engineering Supervisor, MCE D. Jamil, Site Vice President, McGuire Nuclear Station R. Johansen, Standby Shutdown Facility System Engineer J. Lukowski, RATRE Eloctrical Syctoms RES) - Power E. Merritt, RES - Instrumentation and Controls J. Oldham, Fire Protection Engineer, MCE - Civil B. Peele, Station Engineering Manager G. Peterson, Site Vice President, CatawbaMcGuire Nuclear Station C. Thomas, Regulatory Compliance Manager K. Thomas, Manager, RES NRC Personnel J. Brady, Senior Resident Inspector, Shoaron Harris E. DiPaolo, Resident Inspector R. Fanner, Nuclear Safety Intern (Trainee)

C. Ogle, Chief, Engineering Branch Chief1, Division of Reactor Safety, Region II R. Rodriguez, Nuclear Safety Intern (Trainee)

S. Shaeffer, Senior Resident Inspector LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-369,370/03-07-01 URI Fire Suppression System for Altemati'eDedicated Shutdown Areas nNot in Accordance with 10 CFR 50, Appendix R, Section llI.G.3 (Section 1 R05.02.b) 50-369,370/03-07-02 URI Failuro to Protoct Roactor Protoction Systom Cabloe Results in Loss of Roquirodlnadeauate Separation and Attachment

Charles R. Ogie - comparemcq 03

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Paa 3 31 Protection of Cables Associated With Redundant Trains of Instrumentation Located in the Same Fire Area (Section 1 R05.03.b.1) 60 369,37003 07 0350-370/03-07-03 U-R Roqukrments Rolativo to tho Numbor of Spurious Oporation6 that mutt bo Poctulatod (Soction 50 369,370/03-07-04 50 370/03 07 06 4 R05.04.b.4)-

UR Methods for Roactor Coolant Systom Prossuro Contro!

During SSF Oporation (Soction 1 R05.05.b)

U-RI Failuro to Provido Adoquato URI Use of a Local Manual Operator Action in Lieu of Providing Physical Protection for Cables of Redundant Safe Shutdown Equipment in Fire Area 16/18 (Section 1 R05.03.b.2) 60 370/03 07 0650-369/03-07-03, 370/03-07-04 URI Requirements Relative to the Number of Spurious Operations That Must be Postulated (Section 1 R05.04.b.1) 50-370/03-07-05 URI Spurious Closure of Valve 2CA0007A Could Lead to Damage of the TDAFW Pump (Section 1 R05.04.b.2) 50-369/03-07-04, 370/03-07-06 URI Methods for Reactor Coolant System Pressure.

Control During SSF Operation (Section 1 R05.05.b)

Closed 50-369,370/00-09-04 URI Adequacy of the Fire Rating of Mineral Insulated Cables in Lieu of Thermo-Lag Electrical Raceway Fire Barrier Systems (Section 40A5.01)

Discussed None Attachment

, iCharles R. Ogle - compare-mcg-03 7.wpd Page 32 32 Attachment

i Charles R. Oale - comDare-mca 03-07.wDd Paae 3, k Charles R. _e

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33 APPENDIX LIST OF DOCUMENTS REVIEWED Section 1R05: Fire Protection Procedures APIO/A15500/045, Plant Fire, Rev. 0 and Rev. 2 AP/11A/55001024, Loss of Plant Control Due to Fire or Sabotage, Rev. 21 AP/21A/55001024, Loss of Plant Control Due to Fire or Sabotage, Rev. 20 NSD 112, Fire Brigade Organization, Training, and Responsibilities, Rev. 5 NSD 313, Control of Combustible and Flammable Material, Rev. 4 NSD 314, Hot Work Authorization, Rev. 2 NSD 316, Fire Protection Impairment and Surveillance, Rev. 6 MP/O/A/7650/122, Inspection of Fire Hose and Hydrant Houses, Rev. 5 OP/O/A/6100/020, Operational Guidelines Following a Fire In Aux Bldg or Vital Area, Rev. 16 PT101A/42501004, Fire Barrier Inspection, Rev. 19 PTI0/A/4250101 1, Fire Door Inspections, Rev. 14 PT/OA142501020, Roll-Up Fire Door Semi-Annual Inspection/Test, Rev. 2 PT/0/A/44001001 A, Fire Protection System Periodic Test, Rev. 24 PT/OA/4400/001 C, Fire Protection System Monthly Test, Rev. 54 PT/OA/4400/001 K, Fire Protection Annual Valve Test, Rev. 35 PT/OA/4400/001 M, Fire Protection System Flow Test, Rev. 14 PT/OA/4400/008, Fire Hose Hydrostatic Test SLC-Committed Hose Stations, Rev. 11 PT/OA/4400/01 OA, Main Fire Pump A, Rev. 15 PT/IOA/4400101OB, Main Fire Pump B, Rev. 10 PT/OIA/4400/01 OC, Main Fire Pump C, Rev. 11 PT/O/A/4400/017, Fire Pump A and B Operability Test, Rev. 13 PT/0/A/4400/018, Fire Pump C Operability Test, Rev. 11 PT/I /A/4400/001 L, Fire Protection Containment Header Test, Rev. 9 PT/l/AI4400/001 N, Halon 1301 System Periodic Test, Rev. 29 PT/21A/44001001 L, Fire Protection Containment Header Test, Rev. 7 PT/O/A/4600101 6A, Fire Detection System Operational Tests, Rev. 18 PT/O/B/4600/015, Fire Detection System Monthly Test, Rev. 14 PT/01/A47001049, SLC Fire Hose Inspection, Rev. 1 PT/i/At4700/042, SLC Fire Hose Station Valve Operability Test, Rev. 3 PT/2/A47001043, SLC Fire Hose Station Valve Operability Test, Rev. 3 PT/1 /A/41501001 B, Reactor Coolant Leakage Calculation, Rev. 47 Drawings MC-1042-4, General Arrangement, Auxiliary Building, Elevation 750+0, Rev. 6 MC-1201-2-A, General Arrangement, Auxiliary Building, Elevation 716+0, Rev. 67 MC-1201-3-A, General Arrangement, Auxiliary Building, Elevation 716+0, Rev. 67 MC-1201-4, General Arrangement, Auxiliary Building, Elevation 733+0, Rev. 27 Attachment

A Charles R. Ogle - compare-mc2-O3-O7.wpd Charles R. 0 le - corn are mc 03 07.w d Page 34 34 MC-1223-38, Auxiliary Building, Unit 1 & Unit 2, Beam Schedule at Elevation 733+0, Concrete and Reinforcing, Sheet 1, Rev. 4 MC-1223-39, Auxiliary Building, Unit 1 & Unit 2, Beam Schedule at Elevation 733+0, Concrete and Reinforcing Sheet 2, Rev. 6 MC-1223-6, Auxiliary Building, Unit 1, Plan at Elevation 733+0, Reinforcing Sheet 1, Rev. 8 MC-1 223-7, Auxiliary Building, Unit 2, Plan at Elevation 733+0, Reinforcing Sheet 2, Rev. 5 MC-1223-8, Auxiliary Building, Unit 1, Plan at Elevation 733+0, Reinforcing Sheet 3, Rev. 6 MC-1223-9, Auxiliary Building, Unit 2, Plan at Elevation 733+0, Reinforcing Sheet 4, Rev. 6 MC-1223-27, Auxiliary Building, Units 1 & 2, Sections at Elevation 733+0, Concrete Sheet 3-1, Rev. 27 MC-1224-9, Auxiliary Building Unit 1, Plan at Elevation 750+0, Reinforcing Sheet 3, Rev. 9 MC-1 224-1 0, Auxiliary Building Unit 1, Plan at Elevation 750+0, Reinforcing Sheet 4, Rev. 10 MC-1224-39, Auxiliary Building, Beam Schedule at Elevation 750+0, Concrete & Reinforcing Sheet 1, Rev. 6 MC-1 225-1 0, Auxiliary Building Unit 2, Plan at Elevation 767+0, Reinforcing Sheet 4, Rev. 5 MC-1 225-11, Auxiliary Building, Plan at Elevation 767+0, Reinforcing Sheet 5, Rev. 4 MC-1225-39, Auxiliary Building, Beam Schedule at Elevation 767+0, Concrete & Reinforcing, Rev. 6 MC-1225-40, Auxiliary Building, Beam Schedule at Elevation 767+0, Concrete & Reinforcing, Sheet 2, Rev. 5 MC-1226-8, Auxiliary Building, Plan at Elevation 784+0, Reinforcing Sheet 3, Rev. 1 MC-1226-9, Auxiliary Building, Plan at Elevation 784+0, Reinforcing Sheet 4, Rev. 2 MC-1226-19, Auxiliary Building, Beam Schedule at Elevation 784+0, Concrete and Reinforcing, Rev. 1 MC-1 315-01.02-105, General Arrangement, Fire, Flood & HVAC Boundaries, Elevation 716+0, Rev. 0 MC-1384-06.02, Fire Protection Layout, Plan at Elevation 716+0, Rev. 7 MC-1384-06.03, Fire Protection Layout, Plan at Elevation 733+0, Rev. 7 MC-1384-06.04, Fire Protection Layout, Plan at Elevation 750+0, Rev. 7 MC-1384-06.05, Fire Protection Layout, Plan at Elevation 767+0, Rev. 7 MC-1384-07.12-00, Fire Plan, Auxiliary Building, Elevation 695+0, Rev. 3 MC-1384-07.01-00, Fire Plan, Unit 1 Turbine Building, Elevation 739+0, Rev. 11 MC-1384-07.13-00, Fire Plan, Auxiliary Building, Elevation 716+0, Rev. 12 MC-1384-07.13-01, Fire Plan, Auxiliary Building, Elevation 71 6+0, Rev. 9 MC-1384-07.14-00, Fire Plan, Auxiliary Building, Elevation 733+0, Rev. 12 MC-1384-07.14-01, Fire Plan, Auxiliary Building, Elevation 733+0, Rev. 9 MC-1384-07.14-02, Fire Plan, Auxiliary Building, Elevation 733+0 & 736+6, Rev. 9 MC-1384-07.14-03, Fire Plan, Auxiliary Building, Elevation 733+0 & 736+6, Rev. 9 MC-1384-07.15-00, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 10 MC-1384-07.15-01, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 2 MC-1384-07.15-01, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 3 MC-1 384-07.15-01, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 9 MC-1384-07.15-02, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 10 MC-1384-07.16-00, Fire Plan, Auxiliary Building, Elevation 760+6, Rev. 7 MC-1384-07.17-00, Fire Plan, Auxiliary Building, Elevation 767+0, Rev. 10 MC-1384-07.17-01, Fire Plan, Auxiliary Building, Elevation 767+0, Rev. 9 MC-1 384-07.18-01, Fire Plan, Auxiliary Building, Elevation 778+10, Rev. 8 Attachment

, Charles R. Ogle - c mparemc2_O3_O7.wpd Page 35 35 MC-1518-06.43-00, Piping Layout, Interior Fire Protection, Nuclear Service Water Pumps, Sprinkler Addition, Rev. 1 MC-1518-06.43-01, Piping Layout, Interior Fire Protection, Component Cooling Pumps, Sprinkler Addition, Rev. 1 MC-1518-25.85-01, Piping Layout, Service Water Piping, Outside Pumphouse, Rev. 29 MC-1 710-01.00, Plan, Control Room Computer Room, Elevation 767+0, Rev. 49 MC-1710-04.08, Battery Room Junction Points Elevation 747, Rev. 15 MC-1710-04.09, Battery Room Junction Points Elevation 746, Rev. 23.

MC-1 710-04.1 0, Battery Room Junction Points Elevation 745, Rev. 20 MC-1 710-04.11, Battery Room Junction Points Elevation 744, Rev. 24 MC-1710-04.12, Battery Room Junction Points Elevation 743, Rev. 22 MC-1710-04.13, Battery Room Junction Points Elevation 742, Rev. 24 MC-1710-04.14, Battery Room Junction Points Elevation 741, Rev. 23 MC-1710-04.15, Battery Room Junction Points Elevation 740, Rev. 23 MC-1762-01.00-02, Location Diagram, Fire Detectors Located on Elevation 716+0, Rev. 7 MC-1762-01.00-03, Location Diagram, Fire Detectors Located on Elevations 733+0 & 739+0, Rev. 10 MC-1762-01.00-04, Location Diagram, Fire Detectors Located on Elevation 750+0, Rev. 10 MC-1762-01.00-06, Location Diagram, Fire Detectors Located on Elevations 760+6 & 767+0, Rev. 13 MC-2901 -01.01, Auxiliary Building Plan Below Elevation 733'+0, Rev. 44 MC-2907-01.01, Penetration and Switchgear Rooms Plan Below Elevation 776'+0, Rev. 25 MCEE-138-00.02, Turbine Driven AFW Suction Supply Valve, Rev. 5 MCEE-1 38-00.04, Turbine-driven AFW Suction Supply Valve, Rev. 11 MCEE-1 38-00-01, Turbine Driven AFW Suction Supply Valve, Rev. 5 MCEE-211-00.52, Pressurizer Heaters, Rev. 2 MCEE-211-00.52-01, Pressurizer Heaters, Rev 9 MCEE-211-00.52-02, Pressurizer Heaters, Rev. 8 MCEE-211-00.52-03, Pressurizer Heaters, Rev. 9 MCEE-211-00.52-04, Pressurizer Heaters, Rev. 4 MCEE-211-00.52-05, Pressurizer Heaters, Rev. 3 MCEE-244-02.01, Steam Generator Level and Pressurizer Level, Rev. 4 MCEE-247-1 0.00, Motor Driven AFW Isolation Valve, Rev. 0 MCEE-247-20.00, Turbine Driven AFW Isolation Valve, Rev. 0 MCEE-247-20.01, Turbine Driven AFW Isolation Valve, Rev. 0 MCEE-247-32.00, Turbine-driven AFW Isolation Valve, Rev. 1 MCEE-247-33.00, Turbine Driven AFW Isolation Valve, Rev. OA MCEE-250-00.03, Pressurizer Power-operated Relief Valve MCEE-250-00.03-01, Pressurizer Power-operated Relief Valve MCEE-250-00.06, Pressurizer Power-operated Relief Valve Isolation Valve MCEE-250-00.24, Unit 2 Chemical and Volume Control Isolation Valve, Rev. 01 MCEE-250-00.28, Reactor Vessel Head Vent Valves, Rev. 6 MCEE-250-00.29, Reactor Vessel Head Vent Valves, Rev. 5 MCEE-250-00.33, Reactor Vessel Head Vent Valves, Rev. 5 MCEE-257.00.54, Chemical and Volume Control Containment Isolation Valve, Rev. 3 MCEE-257-00.24, Chemical and Volume Control Containment Isolation Valve, Rev. 5 MCEE-257-00.50, Unit 2 Chemical and Volume Control Isolation Valve, Rev. 6 Attachment

Charles R. Ogle - comparemcg_03_07.wpd Page3E 36 MCEE-257-00.52, Chemical and Volume Control Isolation Valve, Rev. 1 MCEE-257-00.55, Standby Makeup Pump, Rev. 1 MCFD-1 574-01.00, Nuclear Service Water, Rev. 6 MCFD-1 574-01.01, Nuclear Service Water, Rev. 10 MCFD-1599-01.00, P&ID, Flow Diagram of Fire Protection, Rev. 13 MCFD-1599-01.01, P&ID, Flow Diagram of Fire Protection, Rev. 14 MCFD-1599-02.00, P&ID, Flow Diagram of Fire Protection, Rev. 15 MCFD-1599-02.01, P&ID, Flow Diagram of Fire Protection, Rev. 15 MCFD-1599-02.02, P&ID, Flow Diagram of Fire Protection, Rev. 5 MCFD-1599-02.03, P&ID, Flow Diagram of Fire Protection, Rev. 6 MCFD-1599-03.00, P&ID, Flow Diagram of Fire Protection, Rev. 7 MCFD-1599-03.01, P&ID, Flow Diagram of Fire Protection, Rev. 3 MCFD-2554-01.00, Unit 2 Flow Diagram of Chemical and Volume Control System, Rev. 5 MCFD-2554-01.01, Unit 2 Flow Diagram of Chemical and Volume Control System, Rev. 5 MCFD-2554-01.02, Unit 2 Flow Diagram of Chemical and Volume Control System, Rev. 6 MCFD-2554-01.03, Unit 2 Flow Diagram of Chemical and Volume Control System, Rev. 3 MCFD-2554-02.01, Unit 2 Flow Diagram of Chemical and Volume Control System, Rev. 6 MCFD-2554-05.00, Unit 2 Flow Diagram of Chemical and Volume Control System, Rev. 4 MCFD-2574-02.00, Nuclear Service Water, Rev. 12 MCFD-2574-02.01, Nuclear Service Water, Rev. 2 MCFD-2592-01.01, Auxiliary Feedwater System, Rev. 13 MCFD-2592-02.00, Auxiliary Feedwater System, Rev. 2 MCM.1206.07-0074.001, McNeary Insurance Consulting Services, FP-12 MCM.1206.07-0087.001, McNeary Insurance Consulting Services, FP-18 MCSF-1 560.SS-01, Summary Flow Diagram Standby Shutdown System (SSS), Rev. 2 Completed Maintenance And Surveillance Test Procedures/Records Work Order 98410020, PT 2NCLP5151, SSF Pressurizer Level, dated 3/13/02 Work Order 98410021, PT 2NCLP5121 NC Loop D Hot Leg W/R Pressure, dated 3/13/02 Work Order 98410083, PM 2CFLP61 10, S/G D W/R Level, dated 2/28/02 Work Order 98410084, PM 2CFLP61 00, S/G C W/R Level, dated 3/5/02 Work Order 98410085, PM 2CFLP6090, S/G B W/R Level, dated 3/1/02 Work Order 98410086, PM 2CFLP6080, S/G A W/R Level, dated 2/28/02 Cable Installation Data for the Following Components 2CA0007A 2CA009B 2CFLT6080, 6090, 6100, 6110 2NC272AC, 273AC 2NC33A, 35B 2NCLT5151 2NV1012C 2NV842AC 2NV94AC 2NVPU0046 Attachment

I Charles R. Oale - compare mco 03 07.WDd Paae 37 CaLe R. Qle - como'are-'-aO3 O7.wSd Paa 37,_

37 ORN4AC Calculations and Evaluations MCC-1 223.04-00-001 0, Determine the Reactor Coolant Pump Sealwater Flow Requirements for the SSF Auxiliary Makeup Pump, Type II MCC-1223.42-00-0030, Documentation of the Adequacy of the Assured Suction Sources to the CA Pumps, Rev. 8 MCC-1223.49-00-0030, Sprinkler System for Nuclear Service Water Pumps @ Elevation 716-0, Rev. 0 MCC-1 435.00-00-0006, Calculation for the Technical Basis of Fire Barrier Penetration Seals, Rev. 1 MCC-1435.03-00-0002, Fire Exposure to Unprotected Steel Hangers for HVAC Ducts, Rev. 2 MCC-1 435.03-00-0004, Supports for Cable Tray Penetrating Fire Barriers, Rev. 0 MCC-1 435.03-00-0012, MNS Penetration Seal Database and GL 86-10 Evaluations, Rev. 0 MCC-1435.03-00-0013, Fire Protection Code Deviations, Rev. 0 MCS-1435.00-00-0001, Fire Protection Acceptance Specification, Rev. 17 MCS-1435.00.00-0003, Design Specification for Mechanical and Electrical Penetrations; Fire Flood and Pressure Seals National Fire Codes - Volume 1, Codes & Standards: NFPA 13 - Standard for the Installation of Sprinkler Systems, 1978 Edition Design Basis Document MCS-1 223.SS-00-0001, Design Basis Specification for the Standby Shutdown System, Rev. 12 MCS-1465.00-00-0008, Design Basis Specification for Fire Protection, Rev. 4.

MCS-1465.00-00-0022, Design Basis Specification for Appendix R, Rev. 2 Problem Investigation Process Reports Reviewed G-99-00110, McGuire Fire Protection Functional Audit (SITA) SA-99-04(MC)(RA)(FPFA).

M-97-0331 1, All three CA pumps may have been dead headed during the U1 Rx trip recovery.

M-99-01884, GL 86-10 guidance for circuit failure modes, hot short duration, and design basis transients for dedicated shutdown not evaluated for applicability to MNS methodology.

M-99-01886, NFPA code deviations not documented in UFSAR or FHA as per GL 86-10.

M-99-03926, Effect of warmer seal injection water on RCP seals during SSF event not adequately taken into consideration on SMP capacity. Evaluate applicability to McGuire.

M-00-01 900, Unit 1 CA pumps normal suction sources inadvertently isolated following a reactor trip and automatically aligned to RN.

M-00-04466, Evaluate UFSAR Section 9.5-1 Clarifications for Fire Suppression Systems.

M-00-04469, Evaluate Fire Pump Loss Due to Fire in Fire Area 19 and Main Control Room.

M-00-04483, The fire protection RY by-pass lines around 1 RY 113 and 1 RY 114 do not Permit the Maximum Flow for the Largest Sprinkler Demand.

M-00-04487, Fire Brigade Drills Had Not Been Performed Within 10 Years in Areas Considered Safety Significant.

M-00-04491, NRC Appendix R nspection in certain fire areas determined the potential for NC Attachment

I J Charles R. Ogie - comparemc9._03_07.wpd Icharles R. Ogle - comparemcgO~o7.wpd Page 38 38 PORV and block valve actuation. We need to evaluate this cabling as to if" this will occur.

M-00-04516, Adequacy of Pzr heater capacity at SSF due to increase safety valve leakage.

M-02-01708, It has been discovered that pressurizer ambient heat losses are greater than calculated in OSC-3144 impacting SSF ASW system operability (TS 3.10.1 and TS 3.4.9).

M-02-03214, SSS and NC DBDs identified errors related to pressurizer heater requirements.

M-02-05031, RO closed.1 CA-0002, resulted in temp low suction flow to running 1 B CA pump.

M-02-05096, Information on system problem [PIP M-02-05031] not documented for resolution.

M-03-01675, Fire Detection System Not Installed to NFPA Codes.

M-03-01748, Smoldering fire on roof of Unit 1 Diesel Generator building.

Prblem Investigation Process Reports Generated During This Inspection M-03-02084, Fire scenarios that could cause suction loss to U2 TDCA pump for SSF areas.

M-03-02086, Discrepancy between Appendix R DBD and Procedure AP/2/A15500124.

M-03-02091, Unit 1 and Unit 2 HVAC areas do not have fire detectors.

M-03-02092, Discrepancy between drawings and fire pre-plans for fire hose lengths.

M-03-02093, Drawing discrepancy for as-built configuration of HVAC Equipment Room 805A.

M-03-02106, B train cables in A SWGR room Fire Area which are not previously identified.

M-03-02115, Appendix R logic diagrams not updated to show function of valve 2CA002.

M-03021 18, Appendix R logics for AFW do not show valve 2CA0007A.

M-03-02249, Detector zones 203 and 204 not in SLC 16.9.6, Table 16.9.6-1.

M-03-02275, Calculation (MCC 1223.48-00-0030) in support of sprinkler system design over the nuclear service water pumps needs revising.

M-03-02294, SLC Table 16.9.7-1 appears to be missing some information.

M-03-0231 1, Evaluate May 2003 NRC Fire Protection Inspection Items.

M-03-02327, Calc MCC-1 435.03-00-0002 contains deleted pages not marked as being deleted.

M-03-02588, Apparent Appendix R violation in the 1 ETA and 2ETA switchgear HVAC rooms.

Miscellaneous MNS Units 1 and 2 Safety Evaluation Report (SER), March 1978 SER Supplement 2 (SSER 2), Appendix D, Fire Protection Review, Units 1 & 2, March 1979 SSER 5, Appendix B, McGuire SER, Fire Protection Review, Unit 1 & 2 (Revised), April 1981 SSER6, Appendix C, McGuire SER - Standby Shutdown System, February 1983 MNS Updated Final Safety Analysis Report (UFSAR) Section 9.5.1, Fire Protection System UFSAR Section 16.9.7, Selected Licensee Commitments (SLC), Standby Shutdown System Letter from W.O. Parker, Duke Power Co., to H.R. Denton, NRC, McGuire Nuclear Station Fire Protection, dated January 9, 1981 Letter from D.S. Hood, NRC, to H. B. Tucker, Duke Power Co., Fire Protection Deviations, McGuire Nuclear Station, Units 1 and 2, dated May 15, 1989 Fire Area Ventilation Rates, Fire Areas 4, 13, 18 & 24 Fire Area Oil Quantities, Fire Area 4, 13, 18 & 24 Fire Area 4 Correlation List between Rooms Number vs. Detection Zones Fire Qualification Test on Silicone Foam Floor Pen Seals, Slab No. 5, Project No. 03-5656-001 Applicable Codes and Standards Attachment

,i Charles R. Ogle - compare_mcg_03_07.wpd It*

Page 3S 39 NFPA 13, Standard for the Installation of Sprinkler Systems, 1978 Edition NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1976 Edition NFPA 72E, Standard on Automatic Fire Detectors, 1974 Edition Modifications Minor Modification MM-12907A thru F Attachment

Charles R. 0q1e - compare-mcq-03-07.wpd Paae 40 Charl s R. Ogle-r mcaa 4

40 LIST OF ACRONYMS AMU Air Handling Unit AARA A _ I _... ^ _ ^-..

AE LO#V AE liGaEOR8hlV AchGnUnhlb raw ANS

-Amorican Nucloar Standard ANSI Amorican National Standards Instituto A E -A k- -

I A 0 R A A Radiation Monitor ASME Amorican SocioW-C--;

AL"r1fi A--.:-.

IUI IVIUbIIdIIItl CllUIIIUUlt, P=7 1 iVi Ill8iIlCt^{

Rcinintu Im Tartinn Untarin'r, 7sB 7 1 IV>1 GA-ABAuxiliary Building r-ormrUixt; Artign PFGgFaFn GGF AFW Auxiliary Feedwater-GAP YEA w

-kAdRffal loc-l"Effa!'UR P-cWHRY k71-Feadwa"H-Af-l w

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GFR Geele e; FedeFal Regulavons Ge Gebait-6W 6AWA164y FFGG9GlUF8 WRG Duke FleweF Company-DRP-D1sGFete RadieaGtiVe RaFtiGle SGGS-6meFqenGy GeFe Geeling System SD E-19GUORiG DasimeteF El)G-SFAeF9eRGy Diesel GeReFateF 614IF-Effluent MeRitgFiR
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EnRad ERViFgFimental Radiation SGG-E-Rd Of GyGle EP EMeFgeRgy PFOGedUF9 ESF-EngineeFed 8afeguaFdr,_

FeatUFe ESPAS E-F;qiF;9eFed Safety Featum AGtuaUei+-

System P=-Vital BatteFy G RNST-RefueliFig Water-Steh-age TaFik GPM-GalleRs PeF Minute GV-GeyemoF-Valve GWR Ga6eeus Waste Relea6e HP-Health Ph HRA-High RadiatiOR AFea 149PA High 64i '

PaFtiGulate AiF IN120-IRstitute of NuGleaF PeweF OpWat' AP Abnormal Procedure DSD Dedicated Shutdown FHA Fire Hazards Analysis FPP Fire Protection Review GL Generic Letter HVAC Heating Ventilation and Air Conditioning IPEEE Individual Plant Examination for External Events IR Inspection Report

!SFS!

lRd9pei;d9;;t Spent FU91 StWage iRstallatio I CO Limiting Genditien fGF OPMUGR LER-LIGGA699 EveRt RepeFt LHRA-LeGked High R;;di;;tinn ArAR I I D-1 nwA I Unit Ai DA Action I=QGA Less of Geelant.Uniden I UR Liquid WasW Mfd0Udbfd FAkM 0 M 0 UFfjPFdFdFY MNS-kWKilowatt MCR Main Control Room MNS McGuire Nuclear Station KGNC Geoling water ISIG-19 190-RA n

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-Residual Heat Remeval NSI-NuGleaF 9;;eF
y !Rstitute NI Safety injeGfien NQED NetiGe Of Snf0FG9FA9At Attachment

Charles R. Oale - comDare-mca-03-07.wDd Paae 41 Chales R.

e - co Dar 41 Discrotion NSD Nucloar Site Diroctivo-NV-Reactor Coolant NFPA NRC NRR NSD NV GQ6M National Fire Protection Association Nuclear Regulatory Commission NRC Office of Nuclear Reactor Regulation Nuclear System Directive Chemical and Volume Control OffnitA Dono Clriltion M, mua

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-Post Accidont Uas Samp ng Systom P1 Forformanco Indicater PlP-Problem Investigation Process Testing PS Public roport PMT l

Rt-A:M ifltofal Aco RAdiaion Safty PT

-Porformanco Toct PWR wrnr.r.llri7nn Wntnr IRnnrtnr QCQuality Control AB -Roactor Auxiliary Building RAP Regulated Air Pump RCA -adiologically Controllod Aroa RCZ -Radiation Control Zono RD Radiation Doimetr; and Rec^rds Procoduro iFMW inaininnirnl Fnvirnnmnntnl R A--!4-.:--

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c t o Sy.s 12RPSReactor Protection System SDP Significance Determination Process SPF Significant Event Invoctigation Toam SFP Spen Fel-Pool SH Sharod Hoafth Physics Proceduro SLC SERSafety Evaluation Report SG SLC SSG SSA SSD SSF SSRS

• TEDE Steam Generator Selected Licensee Commitment

-SMP Structuro., Systome, ComponontsStandby Maket Safe Shutdown Analysis Safe Shutdown Standby Shutdown Facility Solid Stato ProtectionStandby Shutdown System

-TDAFW Turbine-Driven Auxiliary Feedwater Total ffoctivo Doso Eauivalont TH up Pump L tl.46L l umpvrary nu CH"I.-

Phycice Procoduro Tl 4e rperary- -

Instruction TLD Thormoluminoccont Attachment

i Charles R. Ole - compare-mcg_03_07.wpd Page 42 0

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42 DOsleteF-TS-Technical Specifications JJ~I4U-2-FSAR -

Updated Final Safety Analysis Report Volumo Control Tank WBC Wholo body Count WGDT Wasto Gas oray Tank WO Work Ordor YC Chillod Wator (GotrIo Foom)URIUnresolved Item Volt V

Attachment

V)00001.TMP

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MY COMMENTS ON MCG IR 03-07 7/8/03 11:06AM Charles R. Ogle CRO@nrc.gov Recipients nrc.gov AT1_PO.ATL_DO DCP CC (Charlie Payne)

MXT2 (McKenzie Thomas)

PJF (Paul Fillion)

Action Delivered Opened Opened Opened Date & Time 07/08/03 11:06AM 07/09/03 09:42AM 07/08/03 11:45AM 07/08/03 01:03PM Post Office AThPO.ATL_DO Delivered 07/08/03 11:06AM Route nrc.gov Files commentsMcgO3_07.wpd comparecmcg_03_07.wpd MESSAGE Options Auto Delete:

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