ML042380302

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E-mail, C. Ogle to Charlie Payne, My New Comments on Bf IR 03-07
ML042380302
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 12/01/2003
From: Ogle C
NRC/RGN-II
To: Payne D
NRC/RGN-II
References
FOIA/PA-2004-0277 IR-03-007
Download: ML042380302 (37)
v • d • e


Text

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-I-- --- -100?Aj From: Charles R. Ogle / Cd, To: Payne, Charlie Date: 12/1/03 2:30PM

Subject:

my review comments on BF IR03-07 I thought you did an excellent job on this Inspection report. I previously gave you my comments on what I believe was rev 2. Attached Is a comparison between Rev 3 and what was sent out.

General comments:

1.Cover letter and body of report. Please try to be consistent on the numbering/designation of areas or things described Inthe report. It makes it more difficult for the reader Ifthere are even subtle name changes Inthese Items as the report progresses.

2. The revised bullet in the summary section for the first finding does a much better Job of describing why the finding isminor.
3. Section 1R05.01 .b. Enforcement: The revised version Is more to the point and eliminates a lot of regulatory discussion which Istangential to the issue. I'd like to use this direct regulatory approach Inthe future.

REO4 UNITED STATES l zNUCLEAR REGULATORY COMMISSION REGION II SAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET SW SUITE 23IT5 ATLANTA, GEORGIA 303034931 November 17, 2003 Tennessee Valley Authority ATTN: Mr. J. A. Scalice Chief Nuclear Officer and Executive Vice President 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000260/2003007 AND 05000296/2003007

Dear Mr. Scalice:

On October 3, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an Inspection at your Browns Ferry Nuclear Plant Units 2 and 3. The enclosed Inspection report documents the inspection findings, which were discussed on that date with Mr. A. Bhatnagar and other members of your staff. Following completion of additional review in the Region II office, a final exit was held by telephone with Mr. J. Lewis and other members of your staff on November- 147, 2003.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.

The Inspectors reviewed selected procedures and records, observed activities, and Interviewed personnel.

This report documents one finding concerning a failure to protect certai, mn. safety _ntrol eireuit cables and instead using unapproved,procedural guidance directing a local manual operator actions in the Unit 3 480 Yvolt Reactor Motor Operated Valve Board Room-BARoom 3A during a severe fire in that location. This finding has potential safety significance greater than very low significance. This finding did not present an Immediate safety concem. In addition, the report documents one NRC-identified finding of very low safety significance (Green) Involving a violation of NRC requirements. However, because of the very low safety significance and because It is entered into your corrective action program, the NRC is treating this finding as a non-cited violation (NCV) consistent with Section VL.A of the NRC Enforcement Policy. If you contest any NCV Inthis report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator Region 2; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Browns Ferry Nuclear Plant.

I

TVA 2 In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection In the NRC Public Document Room or from the Publically Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at httQ:/1Www.nrc.aov/

readina-rm/adams.html (the Public Electronic Reading Room).

Sincerely, IIRA Charles R. Ogle, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-260, 50-296 License Nos.: DPR-52, DPR-68 Endlosur : (See page 2)

Enclosure:

NRC Triennial Fire Protection Inspection Report 05000260/2003007 and 05000296/2003007 w/

Attachment:

Supplemental Information cc w/encl:

Karl W. Singer Senior Vice President Nuclear Operations Tennessee Valley Authority Electronic Mail Distribution James E. Maddox, Vice President Engineering and Technical Services Tennessee Valley Authority Electronic Mail Distribution Ashok S. Bhatnagar Site Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution General Counsel Tennessee Valley Authority Electronic Mail Distribution Michael J. Fecht, Acting General Manager Nuclear Assurance Tennessee Valley Authority Electronic Mail Distribution

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

(cc w/encl cont'd)

Michael D. Skaggs, Plant Manager Browns Ferry Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution Mark J. Burzynski, Manager Nuclear Licensing Tennessee Valley Authority Electronic Mail Distribution Timothy E. Abney, Manager Licensing and Industry Affairs Browns Ferry Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution (cc vdeeln cont'd See page 3 (e enel eent~d State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P. O. Box 303017 Montgomery, AL 36130-3017 Chairman Limestone County Commission 310 West Washington Street Athens, AL 35611 Distribution wfencl:

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TVA 4 OFFICE RII:DRS RII:DRS RII:DRS CONTRACTOR RII:DRP SIGNATURE A RA RARA NAME SWALKER GWISEMAN CPAYNEKS UWVA S CAHILL lDATE 11/10/2003 11/17/2003 11/17/2003 1 11/17/2003 E-MAIL COPY? YES N YES NO YES NO YES NO YES NO Y YES NO PUBUC DOCUMENT YES NO I I I I I relfA-ZA~~~~~~~~~~~

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U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos.: 50-260, 50-296 License Nos.: DPR-52, DPR-68 Report No.: 05000260/2003007 and 0500029612003007 Licensee: Tennessee Valley Authority Facility: Browns Ferry Nuclear Plant Location: Corner of Shaw and Nuclear Plant Roads Athens, AL 35611 Dates: September 8-12, 2003 (Week 1)

September 29 - October 3, 2003 (Week 2)

Inspectors: C. Payne, Senior Reactor Inspector (Lead Inspector)

S. Walker, Reactor Inspector G. Wiseman, Fire Protection Inspector K. Sullivan, Consultant, Brookhaven National Laboratory Accompanying N. Staples, Nuclear Safety Intern Personnel: R. Rodriquez, Nuclear Safety Intem Approved by: Charles R.Ogle, Chief Engineering Branch I Division of Reactor Safety Enclosure

SUMMARY

OF FINDINGS l IR 0500026012003-007, 05000296/2003-007; 9/8 - 1£003911 2/2003 and 9/29 - 10/3/2003; Browns Ferry Nuclear Plant, Units 2 and 3; Triennial Fire Protection.

The report covered an announced two-week period of Inspection by three regional Inspectors and a consultant from Brookhaven National Laboratory. One Green non-cited violation (NCV) and one unresolved Item 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. NRC-Identified and Self-Revealing Findings Comerstone: Mitigating Systems

  • TBD. The inspectors identified a violation having potential safety significance greater than very low significance because a fire Safe Shutdown Instruction for Fire Area 13 (Unit 3 480 volt Reactor Motor Operated Valve Board Room 3A) directed an operator to enter the lieensee failed to protect the non safety control circuit cables forlocation of the fire to perform a local manual action associated with tripping the Unit 3 Reactor Recirculation Pumps (RRPs) in Fire Area 13 (Unit 3 400 V Reactor Motor Operated Valve Board Rfoom SA). In lieu of protecting these eables, the liensee used manuhl operator actionls to leeally trip the Unit 3 Rlrvs but failed to obtain prior NR approval. Additionally, -during a severe frbe in Fire Area 13, these mantial actions would be performed in Flre Area 4and.

I This action may not be successful for a severe fire In this room because of the high temperatures, heavy smoke, low visibility and hazardous plant conditions that would likely be encountered by the operator while the actions are performed.

This finding Is unresolved pending completion of a significance determination.

This finding Is greater than minor because It adversely impaeted the capability of systems, structures and components necessary to achieve and maintain the plant in a safe shutdown eonditon during a severe fire and effeetedis associated with procedure quality and degraded the reactor safety mitigating systems cornerstone objective. The finding was determined to have potential safety significance greater than very low significance because If the RRPs are not tripped, the RRP discharge head pressure could impede Residual Heat Removal (RHR) Low Pressure Coolant Injection (LPCI) flow. RHR LPCI flow is the assured method for maintaining reactor water level in the safe range during severe plant fires. Inadequate RHR LPCI flow may cause reactor core uncovery and potential fuel damage. (Section 1R05.01)

2 the requirement to Implement fire watches for Impaired fire protection systems and features. On October 23, 2002, the licensee Inappropriately used the fire protection license change process to revise the FPP to permit the removal of fire suppression systems and/or fire rated barrier assemblies, necessary to satisfy the separation and suppression requirements of 10 CFR 50, Appendix R, Section III.G, from service without compensatory measures being Implemented (i.e., fire watches being posted) In the affected plant areas. The change could adversely affect the ability to achieve and maintain safe shutdown (SSD) in the event of a severe fire in the affected area.

This issue was not assessed in accordance with the SDP but instead was assessed in accordance with guidance in Sections IV.A.1 through IV.A.4 and Section IV.B of the NRC's Enforcement Policy. The Issue was significant because the licensee's change process for the FPP allowed this degraded condition to be accepted without prior NRC approval. The inspectors concluded that this issue had a credible impact on safety because the licensee's failure to properly evaluate the removal of fire watch posting requirements could adversely affect or degrade the ability for achieving and maintaining SSD from the main control room, local shutdown stations, or alternate shutdown stations. However, the inspectors determined that this finding was of very low significance because, based on an assessment of the impacts of the Identified fire protection features removed from service, the licensee's overall SSD capabilities in the affected fire areas and related FPP features (fire brigade) remained adequate to achieve and maintain SSD conditions. Therefore, this finding is characterized as Green.

(Section 1R05.11 )

B. Licensee-identified Violations None

REPORT DETAILS

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity 1R05 Fire Protection The purpose of this inspection was to review the Browns Ferry Nuclear Plant 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 SSD systems Is maintained free of fire damage. The Inspection was performed in accordance with the U.S. Nuclear Regulatory Commission's (NRC) Reactor Oversight Process using a risk-informed approach for selecting the fire areas and attributes to be inspected. The inspectors used the licensee's Individual Plant Examination for External Events and in-plant tours to choose three risk-significant fire areas for detailed inspection and review.

The fire areas chosen for review during this Inspection were:

  • Fire Area 9, Unit 2 4 kilovolt (kV) Shutdown Board Room C and 250 volt (V)

Battery Room, Unit 2 Reactor Building, 621 foot (ft.) level.

  • Fire Area 13, Unit 3 -480 V Reactor Motor Operated Valve (RMOV) Board Room 3A, Unit 3 Reactor Building, 621 ft. level.
  • Fire Area 3-4, Unit 3 Reactor Building, 621 ft. and 639 ft. level.

The inspectors evaluated the licensee's FPP against applicable requirements, including Operating License Conditions 2.C.(14) [Unit 2] and 2.C.(7) [Unit 3]-,FPP; Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), Appendix R; 10 CFR 50.48; Branch Technical Position (BTP) Chemical and Material Engineering Branch (CMEB) 9.5-1; related NRC safety evaluation reports (SERs); the Browns Ferry Nuclear Plant Updated Final Safety Analysis Report (UFSAR); and plant Technical Specifications (TS). The inspectors evaluated all areas of this Inspection, as documented below, against these requirements.

Documents reviewed by the inspectors are listed in the attachment.

.01 Systems Reauired to Achieve and Maintain Post-fire Safe Shutdown

a. Inspection Scope The licensee's Safe Shutdown Analysis Report (SSAR) was reviewed to determine the components and systems necessary to achieve and maintain SSD conditions in the event of fire Ineach of the selected fire areas. The objectives of this evaluation were:
  • Verify that the licensee's shutdown methodology has correctly identified the components and systems necessary to achieve and maintain an SSD condition.

2

  • Confirm the adequacy of the systems selected for reactivity control, reactor coolant makeup, reactor heat removal, process monitoring and support system functions.
  • Verify that an SSD can be achieved and maintained without off-site power, when it can be confirmed that a postulated fire in any of the selected fire areas could cause the loss of off-site power.

The main control room (remote) and In-plant manual operator actions (local) for controlling plant operation, fire response and achieving a SSD condition in response to a severe plant fire were reviewed and walked down by the inspectors. The inspectors evaluated the following plant procedures to accomplish this task.

  • 0-AOI-26-1, Fire Response
  • 2-AOl-100-1, Reactor Scram
  • 2/3-SSI-001, SSD Instructions
  • 2/3-SSI-3-4, Unit 3 Reactor Building Fire Elevation 621 & Elevation 639 North of R-Line
  • 2/3-SSI-9, Unit 2 Reactor Building Fire 4 kV Electric Board Room 2A
  • 2/3-SSI-13, Unit 3 480 V Reaeter Motor Operated Valve (RMOVt Board Room 3A
  • EPIP-17, Fire Emergency Procedure
b. Findings

Introduction:

A finding was identified in that the lieensee failed to protect the non safety Io ireuit atrfl eables ferSafe Shutdown Instruction (SSI) for Fire Area 13 (Unit 3 480 V RMOV Board Room 3A) directed an operator to enter the location of the fire to perform a local manual action associated with tripping the Unit 3 Reactor Recirculation Pumps (RRPs) in Fire Area 13 (Unit 3400 ' Reaetr Moztor Operated i alve BCard Room 3A).

In lieu of protecting these cables, the licensee used manual operator actions to locally trip the Unit 3 .. rs but falued t obtain pirior Nfl approval. Additionally, during et severe fire in Fire Area 13, these manual actions would be peItrrned in Fire Area 13 and. This action may not be successful for a severe fire in this room because of the high temperatures, heavy smoke, low visibility and hazardous plant conditions that would likely be encountered by the operator while the eetonseaireaction is performed.

This is an unresolved item (URI) pending completion of the significance determination process (SDP).

Description:

The licensee's SSAR assumes that both RRPs are tripped following a reactor scram. Normally, these actions are accomplished by operators Inthe main control room. However, in certain fire areas, RRP control power eittflscables were not protected from potential fire damage. As a result, the licensee developed manual operator actions to locally trip the RRP recirculation pump trip (RPT) breakers.

During a plant walk down of Procedure 2/3-SSI-13 for a fire in Fire Area 13, the inspectors noted that two steps in Attachment 6 directed the operator to enter the room where the fire was located and locally open an electrical circuit breaker. Specifically,

3 Steps 1.1 and 1.2 of Attachment 6 to 2/3-SSI-13 directed the operator to enter Fire Area 13, go to the 250 V DC Rleaetr Motor Operated Va e,(RMOV) Board 3A (located in Fire Area 13), and place the 4160 V RPT Board 3-1l normal control power breaker to off.

The operator was then to proceed to the reactor building and open the associated RPT breakers (causing the RRPs to trip, if operating). These actions were required to be completed within 20 minutes of initiating procedure 2/3-SSI-13, and likely would be performed while the fire was fully involved and fire brigade response was In progress.

Additionally, no alternative operator guidance was provided In2/3-SSI-1 3 for the situation where the above actions could not be accomplished.-

The inspectors concluded that the high temperatures, heavy smoke, low visibility and hazardous plant conditions during a severe fire would make it unlikely that the operator could successfully accomplish these actions. The inspectors also rcncluded that the RRPl eentrel power e.ireuits were associated non safety eireuits as defined in 10 CFR 507 Appendix R, Seetion Il.C.G2. A review of the Brownrs Ferry CERs found that these manual operater actions had not been approved by the Nzic for use in lieu of protecting the eabes.

Following a self assessment documented in report BFN-OPS-03-009, on July 25, 2003, the licensee identified that Its SSD fire instructions utilized local manual operator actions that had not been evaluated to the performance criteria listed in NRC Inspection Procedure 71111.05, Enclosure 2. The licensee Initiated PER 03-013882-000 to generically review all manual operator actions once final guidance was available from NRC rulemaking as discussed. in NRC letter SECY-03-100. At the time of this inspection, the licensee had not begun any reviews of manual operator action feasibility nor had any problems with specific manual operator actions been identified. This finding Is captured in the licensee's corrective action program (CAP) by this PER.

Analysis: The finding adversely Impacted the eapability of systems, struetures and eernpents necessary to achieve and maintainSSI procedure quality for achieving and maintaining the plant Ina safe shutdown condition during a severe fire. Because the finding affected the reactor safety mitigating system cornerstone objective, the finding is greater than minor. The Inspectors determined the finding had potential safety significance greater than very low significance because if the RRPs are not tripped, the RRP discharge head pressure could Impede Residual Heat Removal (RHR) Low Pressure Coolant Injection (LPCI) flow. RHR LPCI flow is the assured method for maintaining reactor water level in the safe range during severe plant fires. Inadequate RHR LPCI flow may cause reactor core uncovery and potential fuel damage. However, this finding remains unresolved pending completion of a significance determination.

Enforcement: The NC-approved FrP for Browns Ferry Units 2 and 3 commits to 10 CFR 50), Ap~pendix Ri, Zeetion 11.G. Ceetion IILCG.2 states, in part, "wJhere eables and equipment including associated non safety circuits that ecould prevent operation or cause maloperation due to hot shorts, open circuts, or shorts to ground, of redundant trains of systems neceseary to achieve and maintain hot shutdown cenditiens are lecated within the same fire area _utside of primary entaimment, nenof the following means off ensuring that one of the redundant trains is free of fire damage shall be provided: (1) physical protection by a three hour rated fire barrier; (2) physical protection by a separation of more than 20 feet with ne interening combustibles or fire hazards plus fire

4 deteelotn and an autenmatie fire su soro stem, or (3) physloal proteotion by a one hour rated fire barrier plu; fire detection and an automatic fire suppression system."

M~anua! operator potion to respond to rnaloperations is not listed as an eeeeptable method for satisfyin~g this reuiern.

- Unit 3 TS 5.4.1.a requires written procedures be established, Implemented, and maintained covering the activities specified in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Appendix A, Item 6v, requires procedures be properly established for combating plant fires. Contrary to the above, the liensee failed to proteot the nom safety eenti ol eireuit eables for the Unit 3 nrRs whose maloperation during a severe fire eould prevent the redundant train of safe shutdown systems from successfully achieving and maintaining hot shutdown conditions. Inlieu of providing adequate physiosl protetion, the lieensoe used-nmaalfl3-SSI-1 3 was not properly established, in that certain operator actions outside the main control room without obtaining p__r NREC approval. In addition, these manual operator ationspresoribed In this procedure would be performed in the area of the fire. Consequently, required actions to achieve and maintain safe shutdown may not be successfully accomplished.

Pending determination of the finding's safety significance, this finding Is identified as URI 05000296/2003007-01, Failure To Protect Unit 3 Reactor Recirculation Pump Control Circuitry From Flre Dam.age And UnapprovedInadequate Unit 3 Fire Procedure Directs Local Manual Operator Actions Be Performed In Location of Fire.

.02 Fire Protection of Safe Shutdown Capabilitv

a. Inspection Scope For the selected fire areas, the inspectors evaluated the frequency of fires or the potential for fires, the combustible fire load characteristics and potential fire severity, the separation of systems necessary to achieve SSD, and the separation of electrical components and circuits to ensure that at least one SSD path was free of fire damage.

The Inspectors also reviewed the Fire Hazards Analysis (FHA) to verify the fire loading used by the licensee to determine the fire-resistive rating of the fire protection barriers and features. The Inspectors also Inspected the fire protection barriers and features to confirm they were installed in accordance with the codes of record to satisfy the applicable separation and design requirements of 10 CFR 50, Appendix R,Section III.G, and commitments to BTP CMEB 9.5-1. The inspectors reviewed the following documents, which established the controls and practices to prevent fires and to control the storage of permanent and transient combustible materials and ignition sources, to verify that the objectives established by the NRC-approved FPP were satisfied.

  • Fire Protection Report (FPR) Volume 1, Fire Protection Plan
  • FPR Volume 2, Section l-D, Smoking Restrictions

-* TVAN Standard Programs and Processes Procedure SPP-10.10, Control of Transient Combustibles

  • TVAN Standard Programs and Processes Procedure SPP-10.1 1, Control of Ignition Sources (Hot Work)
  • Electrical Preventive Instruction EPI-0-000-MCCO01, Maintenance and Inspection of 480 V AC and 250 V DC Motor Control Centers

5 The inspectors toured the selected fire areas to observe whether the licensee had properly evaluated in-situ combustible fire loads and limited transient fire hazards in a manner consistent with the fire prevention and combustible hazards control procedures.

In addition, the Inspectors reviewed selected weekly fire safety inspection reports and fire brigade response and emergency/incident reports for 2002 and 2003 to assess the effectiveness of the fire prevention program.

The fire brigade Is a dedicated group which is independent of the control room staff.

The inspectors reviewed fire brigade response, fire brigade qualification training, and drill program procedures; fire brigade drill critiques; and drill records for the brigade shifts from January 2001 to May 2003. The reviews were performed to determine whether fire brigade drills had been conducted in high fire risk plant areas and whether fire brigade personnel qualifications, drill response, and performance met the requirements of the licensee's approved FPP.

The inspectors walked down the fire brigade house and examined the response vehicle to assess the condition of fire fighting and smoke control equipment. Fire brigade personal protective equipment was examined to evaluate equipment accessibility and functionality. Additionally, the inspectors observed whether emergency exit lighting was provided for personnel evacuation pathways to the outside exits as Identified Inthe National Fire Protection Association (NFPA) 101, Life Safety Code, and Occupational Safety and Health Administration (OSHA) Part 1910, Occupational Safety and Health Standards. This review also Included examination of whether backup emergency lighting was provided for access pathways to and within the fire brigade house and equipment storage areas Insupport of fire brigade operations should power fail during a fire emergency. The fire brigade self-contained breathing apparatuses were evaluated for adequacy as well as the availability, and refill capability, of supplemental breathing air tanks.

The inspectors reviewed fire fighting pre-fire plans for the selected fire areas to determine if appropriate Information was provided to fire brigade members and plant operators to facilitate suppression of a fire. Team members also walked down the selected fire areas to compare the associated pre-fire plans and drawings with as-built plant conditions. This was done to verify that fire fighting pre-fire plans and drawings were consistent with the fire protection features and potential fire conditions described in the plant FHA.

The inspectors analyzed flow diagrams, circuit wiring diagrams and engineering calculations associated with the heating, ventilation, and air conditioning (HVAC) systems of the 2A and 2B Emergency Battery and Shutdown Board Rooms. This review was done to verify that systems used to place the plant Ina SSD condition would not be Impaired by a battery room fire started as a result of hydrogen gas buildup (from ventilation system problems). The inspectors also reviewed the annunciator response procedure for loss of ventilation Inthe battery rooms to affirm that actions were specified that would ensure hydrogen gas concentrations generated by the station batteries would be maintained below explosive limits. The components and equipment included Inthis review are listed in the attachment.

6 The Inspectors reviewed the licensee's methodology for meeting the requirements of 10 CFR 50.48 and the bases for the NRC's acceptance of this methodology as documented In NRC SERs. In addition, the Inspectors reviewed plant documentation, such as the UFSAR, submittals made to the NRC by the licensee Insupport of the NRC's review of their FPP, and deviations from NRC regulations to verify that the licensee met license commitments. Additionally, design control procedures were reviewed to verify that plant changes were adequately reviewed for the potential impact on the FPP, SSD equipment, and procedures as required by the Browns Ferry Unit 2 and Unit 3 operating license conditions. The Inspectors reviewed the criteria Inplant procedures SPP-7.1, On Line Work Management, and SPP-9.3, Plant Modifications and Engineering Change Control, to determine if risk significant plant modifications were developed, reviewed, and approved per the procedure requirements.

b. Findings No findings of significance were identified.

.03 Post-fire Safe Shutdown Circuit Analysis

a. Inspection Scope The Inspectors reviewed how systems would be used to achieve and maintain reactivity control, over-pressure protection, inventory control with high or low pressure injection systems, and residual heat removal during and following a postulated fire in the areas selected for Inspection. The inspection specifically focused on the minimum required systems and equipment necessary to achieve and maintain hot shutdown conditions because damage to these systems could pose a significantly greater risk than damage to systems required to achieve cold shutdown conditions. in additm, the reviewed a sample of the I !VAG system for the selected fire areas.

Portions of the licensee's Appendix R Report which described the methodology and corresponding system flow diagrams were reviewed. Control circuit schematics were analyzed to Identify and evaluate cables important for achieving a SSD. The inspectors traced the routing of cables through fire areas selected for review by using cable schedules and separation calculations and analyses. The inspectors walked down these fire areas to compare the actual plant configuration to the layout indicated on the drawings. The Inspectors also utilized this Information to determine if the requirements of Section lll.G to 10 CFR 50, Appendix R (for protection of control and power cables) were met. The components and equipment Included in the review are listed in the attachment.

b. Findings No findings of significance were identified.

.04 Altemative Shutdown Capabilitv

a. lnspection Scone

7 The selected fire areas that were the focus of this inspection all Involved a reactor shutdown from the main control room. None Involved abandoning the main control room and using an alternative method for achieving SSD from outside of the main control room. The previous NRC triennial Inspection had reviewed this area with no findings of significance (NRC Inspection Report 259,260,296/2000-008). No changes were made to the alternative shutdown methodology Inthe intervening period. Thus, alternative shutdown capability was not reviewed during this Inspection.

b. Findings No findings of significance were identified.

.05 Operational Implementation of Alternative Shutdown Capability

a. Inspection Scone The selected fire areas that were the focus of this Inspection all Involved a reactor shutdown from'the main control room. None Involved abandoning the main control room and using an alternative method for achieving SSD from outside of the main control room. The previous NRC triennial Inspection had reviewed this area with no findings of significance (NRC Inspection Report 259,260,296/2000-008). No changes were made to the alternative shutdown methodology in the Intervening period. Thus, alternative shutdown capability was not reviewed during this Inspection.
b. Findings No findings of significance were Identified.

.06 Communications

a. Inspection Scope The Inspectors reviewed plant communication capabilities to evaluate the availability of the communication systems to support plant personnel in the performance of manual operator actions for shutdown, fire event notification, and fire brigade fire fighting duties.

The Inspectors reviewed the licensee's communications systems' separation analysis to verify -that site portable radio and sound-powered phone systems were designed eonsistent wither the licensing basis and would be available during fire response activities. The Inspectors also reviewed the fire brigade drill critiques to assess proper operation and effectiveness of the fire brigade command post radio communications. In addition, the inspectors reviewed the fire brigade radio communications systems to assess whether the radio channel features would continue to operate if the radio repeaters for the primary communications system became unavailable.

b. Findings No findings of significance were identified.

8

.07 Emeraencv Lighting

a. Inspection Scope The inspectors reviewed the design, operation, and manufacturers data sheets for, the direct current (DC) self-contained battery powered emergency lighting units (ELUs).

The inspectors evaluated the capability of the ELUs to support plant personnel In the performance of SSD functions, Including local manual operator actions, and for illuminating access and egress routes to the areas where those manual actions would be performed. The inspectors checked that these battery power supplies were rated with at least an 8-hour capacity, as required by Section llI.J of 10 CFR 50, Appendix R.

During Inspector walk downs of the plant areas where operators performed local manual actions, the Inspectors inspected area ELUs for proper operation and checked the aiming of lamp heads to determine if sufficient illumination would be available to adequately Illuminate the SSD equipment, the equipment identification tags, and the access and egress routes thereto, so that operators would be able to perform the actions without needing to use flashlights. The inspectors also reviewed completed surveillance and maintenance procedures and test records to ensure that the licensee properly maintained the lighting equipment.

b. Findings I.. No findings of significance were identified.

9

.08 Cold Shutdown Repairs

.a Inspection Scone The licensee had Identified no need fol icensee's SSAR did not identify a need for post-fire repairs to achieve a cold shutdown condition. Thus, cold shutdown repairs were not reviewed during this Inspection. However, the licensee's analysis relied on post-fire ventilation system realignment to remove smoke from fire-affected and surrounding areas. The Inspectors reviewed the licensee's Appendix R emergency ventilation procedure for smoke removal [Abnormal Operating Instruction (AOI) 0-AOI-26-1, Fire Response] and Inspected the portable equipment and ventilation ducts stored at a special Appendix R storage area and other locations onsite used for cooling components In electrical equipment rooms that were required for achieving a cold shutdown condition.

.b Findings No findings of significance were identified.

10

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

a. Inspection Scope The inspectors 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 to ensure that at least one train of SSD equipment would be maintained free of fire damage. The Inspectors selected several fire barrier features listed in the attachment for detailed evaluation and inspection to verify proper Installation and qualification. The Inspectors walked down the selected fire areas to observe the material condition and configuration of the Installed fire barrier features, as well as, reviewed 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 inspectors also reviewed the FHA to verify the fire loading used by the licensee to determine the fire resistance rating of the fire barrier enclosures. The Inspectors also compared the penetration seal ratings with the ratings of the barrier enclosures Inwhich they were installed. The inspectors reviewed the installation instructions for fire doors, the design details for mechanical and electrical penetrations, the penetration seal database, generic ILetter (GL) 86-10 evaluations, and the fire protection penetration seal deviation analysis for the technical basis of fire barrier penetration seals to verify that the fire barrier Installations met design requirements and license commitments. In addition, the inspectors reviewed completed surveillance and maintenance procedures for selected fire barrier features to verify the fire barriers were adequately maintained.

The inspectors reviewed abnormal operating fire procedures, fire fighting pre-plans, fire damper location and detail drawings, and HVAC system drawings for a selected sample of equipment listed in the attachment, to confirm that access to SSD equipment and selected operator manual actions would not be inhibited by smoke migration from one area to adjacent plant areas used to accomplish SSD. Additionally, the inspectors reviewed licensee documentation, such as the UFSAR, submittals made to the NRC by the licensee Insupport of the NRC's review of their FPP, and deviations from NRC regulations to verify that the licensee met license commitments.

b. Findings No findings of significance were Identified.

.10 Fire Protection Systems. Features and Epuipment

a. Inspection Scope The inspectors reviewed flow diagrams, cable routing information, operational valve lineup procedures, and system availability studies, associated with the fire pumps and fire protection water supply system. The Inspectors evaluated the common fire protection water delivery and supply components to determine if they could be damaged or inhibited by fire-induced failures of electrical power supplies or control circuits. Using operating and test procedures, the Inspectors toured the electric motor-driven fire

11 pumps and diesel-driven fire pump to observe the system material condition, consistency of as-built configurations with engineering drawings, and determine correct system controls and valve lineups. Additionally, the inspectors reviewed periodic test procedures for the fire pumps to assess whether the surveillance test program was sufficient to verify proper operation of the fire protection water supply system in accordance with the system operating requirements specified In Sections 9.3 and 9.4 of the FPR.

The inspectors reviewed the adequacy of the design, Installation, and operation of the automatic detection and alarm system for the selected fire areas to actuate in the early stage of a fire. This was accomplished by reviewing engineering drawings for fire detector types, spacing, locations, the licensee's technical evaluation of the detector locations and the ceiling reinforcing plans and beam schedule drawings to determine the location of ceiling bays. After the ceiling bay locations were Identified, the inspectors conducted field tours of the accessible portions of the fire detection systems In Fire Areas 9 and 13 to confirm that detector locations were consistent with the licensee's engineering drawings, FHA, engineering evaluations, and each bay was protected by a fire detector In accordance with the code of record requirements - NFPA 72E, 1990. In addition, the Inspectors reviewed surveillance procedures and the detection system operating requirements specified in Sections 9.3 and 9.4 of the FPR to determine the adequacy of fire detection component testing and to ensure that the detection systems could function when needed.

The Inspectors reviewed the adequacy of the design and installation of the automatic pre-action sprinkler system and water curtains surrounding unsealed vertical openings and the stairwells of the Unit 3 reactor building elevation 621 ft. (Fire Area 3-4). The Inspectors walked down the system to evaluate proper type, placement, spacing of the sprinkler heads, and the extent of the sprinkler head obstructions for effectiveness to prevent a fire from spreading to adjacent fire zones. In addition, the Inspectors examined the sprinkler system hydraulic design calculations to verify that the system could be supplied at sufficient pressure and flow volume to produce the required water density for the protected area. Selected engineering evaluations for NFPA code deviations were reviewed and compared with the physical configuration of the system.

Additionally, the Inspectors reviewed the physical configuration of electrical raceways and SSD components Inthe selected fire areas to determine whether water from a pipe rupture, actuation of the automatic suppression system, or manual fire suppression activities in this area could cause damage that could inhibit the plant's ability to reach a SSD condition.

The inspectors reviewed the manual suppression standpipe and fire hose system to verify adequate design, Installation, and operation in the selected fire areas. The Inspectors examined design flow calculations and flow measurement/pressure test data to verify that the required fire hose water flow for each protected area was available.

The Inspectors performed in-plant walk-downs and observed placement of the fire hoses and extinguishers to confirm consistency with the fire fighting pre-plan drawings.

Additionally, the Inspectors checked a sample of manual fire hose lengths to determine whether they would reach the SSD equipment In the selected fire areas. This was done to ensure that manual fire fighting efforts could be accomplished in the selected areas.

12

b. Findings No findings of significance were identified.

.11 Compensatorv Measures

a. Inspection Scope The Inspectors reviewed the administrative controls for out-of-service, degraded, and/or Inoperable fire protection features, ventilation systems, and post-fire SSD systems and components. The review was performed to verify that the risk associated with removing fire protection and/or post-fire systems or components from service were properly assessed and adequate compensatory measures were implemented In accordance with TS and the approved FPP. The inspectors also reviewed the adequacy of short-term compensatory measures to compensate for a degraded function or feature until appropriate corrective actions were taken.
b. Findings Introduction.: A Green non-cited violation (NCV) was Identified In that the licensee made changes to the approved FPP which decreased the effectiveness of the program without prior Commission approval. The licensee Inappropriately used the License Condition Impact Evaluation (LCIE) change process to revise the FPP to allow the removal of fire suppression systems and/or fire rated barrier assemblies, necessary to satisfy the separation and suppression requirements of 10 CFR 50, Appendix R, Sections III.G.2 and III.G.3, from service without compensatory measures (i.e., fire watches) being implemented.

Description:

The approved Browns Ferry FPP Is documented in the FPR (and incorporated into the UFSAR by reference). The Inspectors reviewed the operating requirements of selected fire protection features specified in Sections 9.3 and 9.4 of the FPR, In SPP-10.9 (Control of Fire Protection Impalmients,) and from the licensee's Fire Protection Impairment Program (FPIP) report logs dated September 29, 2003. During review of the FPIP report log, the inspectors found that the licensee had removed the following fire protection features from service:

  • The pre-action fire suppression sprinkler systems for the Unit 2 and 3 high pressure core injection areas (FPIP No.03-287).
  • Penetration seals in a fire barrier wall separating the Unit 1 reactor building from the control building (FPIP No.03-303).

The inspectors noted that the licensee had removed these fire protection features from service without compensatory measures being Implemented (i.e., without fire watches being posted In the affected plant areas). Upon Investigation, the Inspectors found that the licensee had changed the FPP requirements for Implementing compensatory measures under certain plant conditions.

The Browns Ferry FPP is based on following defense-in-depth (DID) elements (FPR, Volume 1, Section 4.2):

13

  • Prevent fires from starting;
  • Detect fires quickly and rapidly suppress those fires that occur to limit damage; and
  • Design plant safety systems so that a fire which starts in spite of the fire prevention efforts and burns for a significant period of time in spite of fire suppression activities will not prevent essential plant safety functions from being performed.

Defense-in-depth holds that a weakness in one of the above elements can be offset by enhancing the other elements. Fire watches are the most common industry compensatory measure used to help prevent fires. Fire watches strengthen the fire prevention DID element by looking for uncontrolled ignition sources, fire hazards, and combustible materials, and by providing prompt notification of such hazards. In addition, fire watches can strengthen the fire detection and suppression DID element because they are either continuously present within or regularly survey an area for fire. Inthis case, the fire watch would notify the main control room to call out the fire brigade, give the fire brigade exact information about the location and nature of the fire, and may initiate fire suppression activities if trained to do so.

The Inspectors reviewed the licensee's LCIE associated with Revision 20 of the FPR.

This LCIE evaluated changes to the FPR that removed fire watches as a compensatory measure for impairments of the water spray, water sprinkler, or gaseous C02 fire suppression systems and/or fire rated assemblies (i.e., fire barriers). Prior to this change., the Browns Ferry NRC-approved FPP required that, whenever a required fire suppression system and/or fire rated barrier assembly was Inoperable, either a continuous or a one-hour compensatory fire watch patrol (with backup suppression equipment) be stationed. The LCIE concluded that the assignment and presence of fire watch personnel for the purpose of detecting and reporting fires with operable fire detection equipment was unnecessary and provided minimal additional fire protection safety margins. The evaluation also stated that, with the detection system functioning and the alternate suppression equipment available, the response was comparable with the fire watch in place and that the ability to safely shut down the plant was not adversely affected. The inspectors noted, however, that the licensee's change evaluation did not provide a technical basis for these conclusions.

Based on this evaluation, the licensee revised the FPR, Volume 1, Section 9.3.11.C, Spray ancd'or Sprinkler Systems, Section 9.3.1 1.D, CO2 Systemr, oto MA 1.a.3, Fire Rated AssembliesSections 7.5, 9.3.11.C, 9.3.11.D, and 9.3.11.a.3, to delete the requirements for fire watches due to inoperable fire protection systems and features if the associated fire detection system is operable. In addition, rRn Section 7.5 was revised to delete the requirement for a fire watch for areas in which the fire wppreFI;o 1 equipment was inaperabL.

The inspectors concluded that the licensee inappropriately used the fire protection program change process to revise the FPP on October 23, 2002, to permit removing fire suppression systems and/or fire rated barrier assemblies from service without enhancing the other DID elements as a compensatory measure. Specifically, the revised FPP allowed degraded or inoperable fire suppression systems and fire barriers necessary to satisfy the separation and suppression requirements of 10 CFR 50,

14 Appendix R, Sections III.G.2 and IIl.G.3, without establishing compensatory fire watches being-established In the affected plant areas Ifas long as fire detection systems were ftwetioningfunctional. The change adversely affected the ability to achieve and maintain safe shutdown in the event of a fire, Inthat, the licensee went from full compliance with the fire protection safe shutdown system separation and suppression criteria to less than full compliance without implementing eornpensatorytemporary measures to compensate for weakness in this DID element. This was contrary to the safety objectives of the FPP and constituted a change from the approved program that required NRC approval prior to implementation. However, no NRC approval was obtained by the licensee.

AnalVsis.:: Because issues related to the fire protection change process are considered to be findings that could potentially impede or Impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the SDP. In this case, the Issue was significant because the licensee's change process for the fire protection program allowed a decrease in the effectiveness of the fire protection program to be accepted without prior NRC approval. Furthermore, this Issue had a credible impact on safety because the licensee's failure to properly evaluate the removal of fire watch posting requirements could adversely affect or degrade the ability for achieving and maintaining SSD from the main control room, local shutdown stations, or alternate shutdown stations. However, the Inspectors determined that this finding was of very low significance because, based on an assessment of the impacts of the Identified fire protection features removed from service, the licensee's overall SSD capabilities in the affected fire areas and related FPP features (fire brigade) remained adequate to achieve and maintain SSD conditions.

Enforcement. 10 CFR 50.48(a) states, in part, that each operating nuclear power plant must have a fire protection program. Browns Ferry Unit 2 Operating License Condition 2.C.(1 4) and Browns Ferry Unit 3 Operating License Condition 2.C.(7) state, Inpart, that Browns Ferry Nuclear Plant "may make changes to the approved FPP without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain SSD in the event of a fire."

Contrary to the above, the licensee changed the Browns Ferry FPP to remove the requirement to implement fire watches for impaired fire protection systems and features which were a compensatory measure necessary to assure the ability to achieve and maintain safe shutdown in the event of fire. This violation was not evaluated unrihe 6D9 beeause ilmpacted the NRC's ability to perform its regulatory function and, as such, was evaluated in accordance with guidance in Sections IV.A.1 through IV.A.4 and Section IV.B of the NRC's Enforcement Policy. Based on this guidance, this violation of 10 CFR 50.48 and the Unit 2 and Unit 3 Operating License Conditions is classified as a Severity Level IV violation because It resulted in conditions that were evaluated as having very low safety significance. Because this change to the FPP is of very low safety significance and has been entered the finding Into the licensee's CAP (PER 03-018593-000), this violation wasis being treated as an NCV Inaccordance with Section VI.A.1 of the NRC's Enforcement Policy: NCV 05000260,296/2003007-02, Changes Made to the Fire Protection Program Regarding Compensatory Fire Watch Implementation Without NRC Approval.

15

.12 Fire Protection Licensing Basis

a. Inspection Scope The inspectors reviewed licensing basis documents, Including but not limited to SERs and Appendix R exemptions, to ascertain if the Browns Ferry FPP was consistent, and in compliance, with 10 CFR 50.48 and 10 CFR 50, Appendix R. The Inspectors evaluated and compared the licensee's SSD procedures, the FPR, and various calculations of record against the licensing basis to measure the adequacy and consistency of the program documentation.
b. Findings No findings of significance were identified.
4. OTHER ACTIVITIES 40A2 Identification and Resolution of Problems
a. Inspection Scope The inspectors reviewed a sample of licensee audits, self-assessments and PERs to verify that items related to the Browns Ferry FPP, and the capability to successfully achieve and maintain the plant In a SSD condition following a plant fire, were appropriately entered into the licensee's CAP in accordance with the Browns Ferry quality assurance program and procedural requirements. The Items selected were reviewed for classification and appropriateness of the corrective actions taken, or initiated, to resolve the issues. In addition, the Inspectors reviewed the licensee's evaluations of and corrective actions for selected Industry experience Issues related to the fire protection area. The operating experience reports were reviewed to verify that the licensee's review and actions were appropriate. Additionally, the Inspectors reviewed audits and self-assessments of the Browns Ferry FPP to assess the types of findings that were generated and that the findings were appropriately entered Into the licensee's CAP.
b. Findings No findings of significance were Identified.

l 40A6 Meetings. Including Exit On October 3, 2003, the lead inspector presented the Inspection results to Mr. A. Bhatnagar and other members of his staff who acknowledged the findings. The licensee confirmed that proprietary Information was not provided or examined during the inspection. Following completion of additional review in the Region II office, a final exit was held by telephone with Mr. J. Lewis and other members of your staff on November

16 17, 2003, to provide an update on changes to the preliminary Inspection findings. The licensee acknowledged the findings.

SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee personnel:

R. Abbas, Site Engineer Mechanical A. Bhatnagar, Site Vice-President T. Golden, Operations M. Heatherly, Corporate Engineering P. Heck, Site Ucensing Engineer J. Lewis, Operations Manager R. Marks, Site Support Manager R. Rogers, Maintenance Modifications Manager R. Sampson, Site Engineer Electrical M. Skaggs, Plant Manager T. Trask, Design Engineering Manager J. Wallace, Site Licensing Engineer D. White, Nuclear Assurance R. White, Fire Operations Supervisor NRC personnel:

B. Holbrook, Senior Resident Inspector LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 05000296/2003007-01 URI Failure Th PrVteet Unit 3 Rleaetor Recirculation Pump Ccomtrcol Cire itry Fro~m iR.e Damage And UnapprevedInadequate Unit 3 Fire Procedure Directs I Local Manual Operator Actions Be Performed In Location of Fire (Section 1R05.01)

Opened and Closed 05000260,296/2003007-02 NCV Changes Made to the Fire Protection Program Regarding Compensatory Fire Watch Implementation Without NRC Approval (Section lR05.1 1)

Discussed None ATTACHMENT

2 LIST OF COMPONENTS INSPECTED Section 1R05.02: Fire Protection of Safe Shutdown Capabllitv Text Moved i !ere: 1 Component Identification Description 2-AHU-031-2320 Electric Board Room Air Handling Unit 2A 2-AHU-031-2330 Electric Board Room Air Handling Unit 2B 2-FAN-31 -163A 250 V Battery Room Exhaust Fan 2A 2-FAN-31 -163B 250 V Battery Room Exhaust Fan 2B 2-FAN-31 -164A 250 V Battery Room Supply Fan 2A 2-FAN-31-164B 250 V Battery Room Supply Fan 2B 3-FAN-31 -119 Emergency Battery and Shutdown Board Room Exhaust Fan 3A 0-FCO-031 -0093- Emergency Battery and Shutdown Board Room Flow Control Damper End EX MovedA Text I Section I RD5.03: Post-Fire Safe Shutdown Capablity ComDonent Identification Description 0-PMP-026-0001 'A' Electric Fire Pump O-PMP-026-0002 'B' Electric Fire Pump 0-PMP-026-0003 'C' Electric Fire Pump 0-PMP-026-0118 Diesel Fire Pump 2-45N2711-4 ECCS Div. II ATU Inverter 2-FCV-023-0052 RHR Heat Exchanger D Service Water Outlet Valve 2-FCV-067-0021 EECW Sectionalizing Valve 2-FCV-074-0035 RHR Pump 2D Suction Valve 2-FCV-074-0057 RHR System I Isolation Valve 2-FCV-074-0058 RHR System I Containment Spray Isolation Valve 2-FCV-074-0059 RHR System I Suppression Pool Isolation Valve 2-FCV-074-0067 RHR System Ii Inboard Injection Valve 2-FCV-074-0071 RHR System II Isolation Valve 2-FCV-074-0072 RHR System II Containment Spray Isolation Valve 2-FCV-074-0073 RHR System II Containment Spray Isolation Valve 2-FCV-074-0106 RHR Flush Pump Suction Valve 2-PCV-001 -0019 Main Steam Relief Valve 2-PCV-001-0031 Main Steam Relief Valve 2-PCV-001 -0179 Main Steam Relief Valve 2-PMP-074-0039 RHR Pump 2D 2-PNL-9-33 RHR System II Logic Panel PS-3-204AA Main Steam Pressure Switch Div. I ATTACHMENT

3 PS-3-204CB Main Steam Pressure Switch Div. II PS-3-204BA Main Steam Pressure Switch Div. I PS-3-204CA Main Steam Pressure Switch Div. II PS-3-204DA Main Steam Pressure Switch Div. II PS-3-204DB Main Steam Pressure Switch Div. II Section 1R05.09: Fire Barriers and Fire Area/ZonelRoom Penetration Seals Fire Protection Feature Description Fire Barrier Concrete Block Walls North walls of Fire Areas 14 and 15 adjacent to Fire Area 13 Fire Doors Nos. 640, 642, 643, 648, and 654 Fire Dampers Nos. FD-2008, FD-2009, FD-201 0, FD-2577, and FD-2641 Fire Barrier Penetration Seals Nos. S2 6211853, S2 6215071, S2 6215805, S3 6213408, S3 6213467, S3 6215024 Text WAS Moved Fromn 1ler-e: 1 LIST OF DOCUMENTS REVIEWED Procedures 0-AOI-26-1, Fire Response, Rev; 3 0-01-26, Fire Command Center Display (FCCD), Rev. 63 0-SI-4.11.B.1.b, High Pressure Fire Protection System Valve Position Verification, Rev. 35 0-SI-4.11 .B.2.a, Diesel Driven Fire Pump Operability Test, Rev. 29 0-SI-4.11.B.2.C, Diesel Driven Fire Pump Inspection, Rev. 7 0-SI-4.11 .E.1.b(1), Fire Hose Station Operability/Flow Test, Rev. 3 1-ARP-9-20-A, Alarm Response Procedure, Rev. 14 1-ARP-25-1 65, Alarm Response Procedure, Rev. 13 2-AOl-100-1, Reactor Scram, Rev. 75 2/3-SSI-001, Safe Shutdown Instructions, Rev. 5 2/3-SSI-3-4, Unit 3 Reactor Building Fire El. 621 & El. 639 North of R-Line, Rev. 5 2/3-SSI-9, Unit 2 Reactor Building Fire 4 kV Electric Board Room 2A, Rev. 6 2/3-SSI-13, Unit 3 480 V RMOV Board Room 3A, Rev. 5 3-SI-4.11 .C.1.c, Simulated Automatic Actuation of the Fire Protection Sprinkler System, Rev. 22 EPI-0--000-MCC001, Maintenance and Inspection of 480 V AC and 250 V DC Motor Control Centers, Rev. 52 EPIP-17, Fire Emergency Procedure, Rev. 27 TRN-31, Fire Brigade Training, Rev. 5 TVAN FPDP-4, Fire Emergency Response, Rev. 0 TVAN MMDP-1, Maintenance Management System, Rev. 5B1 TVAN SPP-5.4, Chemical Traffic Control, Rev. 2 TVAN SPP-7.1, On Line Work Management, Rev. 4 TVAN SPP-9.3, Plant Modification and Engineering Change Control, Rev. 9 ATTACHMENT

4 TVAN SPP-10.9, Control of Fire Protection Impairments, Rev. 2 TVAN SPP-1 0.1 0, Control of Transient Combustibles, Rev. 2 TVAN SPP-1 0.11, Control of Ignition Sources (Hot Work), Rev. IBI 0-45E643-1, Wiring Diagram, Automatic Fire Detection System, Rev. 10 0-45E724-3, 4160 Shutdown Board C, Rev. 24 0-46E454, Architectural Door and Hardware Schedule, Appendix R, Rev. 5 0-47W216-51, Fire Area Compartmentation and Zone Drawings, Rev. 5 0-47W216-57, Fire Area Compartmentation and Zone Drawings, Rev. 5 0-47W2924-3, Mechanical Heat, Vent, & Air Fire Damper Plans and Sections, Rev. 1 0-47W600-268, Fire Protection System Location Plan, Rev. 0 1-47E859-1, Flow Diagram Emergency Equipment Cooling Water, Rev. 58 2-45C800 Series Drawings, Conduit and Cable Schedule Engineering Safeguards Division I and 11and Engineering Safeguards Division I and 11Reactor MOV Boards - Sh. 2ES-41, Rev. 0, Sh. 2ES-127, Rev. 1, Sh. 2ES-128, Rev. 1, Sh. 2ES-146, Rev. 0, Sh. 2ES-1 57, Rev. 4, Sh. 2ES-1 58, Rev. 1, Sh. 2ES-1 97, Rev. 0 2-45E2750-4, Wiring Diagram 480 V Reactor MOV Board 2B (FCV-74-106) Diagram, Rev. 8 2-45E712-1, 250 V Reactor MOV Board 2A Single Line, Rev. 34 2-45E751-1, 480 V Reactor MOV Board 2A Single Line, Rev. 55

.2-45E751.-2, 480 V Reactor MOV Board 2A Single Line, Rev. 28 2-45E765-4, Wiring Diagram 4160 Shutdown Aux Power (RHR Pump 2D), Rev. 16 2-45E779-22, Wiring Diagram 480 Shutdown Aux Power (FCV-74-67), Rev. 14 2-45E779-47, Wiring Diagram 480 Shutdown Aux Power (FCV-74-35), Rev. 11 2-45E779-49, Wiring Diagram 480 Shutdown Aux Power (FCV-23-52), Rev. 11 2-47E611-74-1, Mechanical Logic Diagram Residual Heat Removal System, Rev. 2 2-47E81 I-1, Flow Diagram Residual Heat Removal System, Rev. 58 2-47EB50, Flow Diagram, Fire Protection and Raw Service Water, Rev. 24 2-47E858-1, Flow Diagram RHR Service Water System, Rev. 18 2-47E2392-630, Penetration Seal Tabular Drawings, Rev. 3 2-47E2865-4, Heating and Ventilation System Control and Relay Rooms Exhaust Air Dampers AOD-HV-I 60-1, -2, Rev. 5 3-45E751-1, 480 V Reactor MOV Board 3A Single Line, Rev. 46 3-45E751-2, 480 V Reactor MOV Board 3A Single Line, Rev. 31 3-45E712-1, 250 V Reactor MOV Board 3A Single Line, Rev. 24 3-45E751-12, 480 V Reactor MOV Board 3E Single Line, Rev. 16 3-47BM491, Mechanical Pre-action Fire Protection Sprinkler System, Reactor Building Subsystem 26-77-El. 621.25, Rev. 0 3-45E643, Wiring Diagram, Automatic Fire Protection System, Rev. 10 3-47E611 1, Mechanical Logic Diagram Main Steam Automatic Depressurization System, Rev. 3 3-47E850-1 0, Flow Diagram, Fire Protection and Raw Service Water, Rev. 10 3-47E2865-4, Flow Diagram, Mechanical Heat, Vent, & Air, Rev. 8 11715-ESK-5BA Motor Driven Fire Pump Supply ACB NAPS Unit 1, Rev. 8 ATTACHMENT

5 11715-ESK-6LC, Elementary Diagram 480 V Circuits Heating and Ventilating: Sh. 11, Rev. 11 11715-ESK-11 F, Eng. Driven Fire Pump 1-FP-P2 NAPS Unit 1: Sh. 1, Rev. 5 11715-ESK-11 F, Eng. Driven Fire Pump 1-FP-P2 NAPS Unit 1: Sh. 2, Rev. 6 12050-ESK-6CK, Elementary Diagram Motor Operated: Sh. 10, Rev. 9 12050-ESK-6LC, Elementary Diagram 480 V Circuits Heating and Ventilating: Sh. 27, Rev. 8 12050-ESK-6PD, Elementary Diagram Solenoid Operated Valves: Sh. 28, Rev. 8 12050-FE-3CF, Wiring Diagram Ventilation Panel Terminal Block Section: Sh. 1, Rev. 20 Calculations. Analyses, and Evaluations 94-0040, Fire Zone Electrical Cable Separation Calculation, Rev. 11 BFN-25-D053 (EPM-R-A-1 11585), Appendix R Fire Pump Availability, Rev. 1 BFN-26-D053 (EPM-ASR-092786), Appendix R Shutdown Board Rooms 2A & 2B Effect of Fire on Embedded Conduits in the Floor Slab, Rev. 0 BFN-ED-NO244-890050, Appendix R Analysis for Intra-plant Communication System, Rev. 3 ATTACHMENT

6 BFN-ND-N0026-920065, PLC, Browns Ferry Nuclear Plant Fire Detection and Alarm System, Detector Selection, Location, and Spacing, Rev. 6 BFN-MD-N0026-910163, Combustible Loads Tables, dated September 29, 2003 BFN-MD-N0031-7030A, 250 V Battery Rooms A, B, C, & D Ventilation Requirements, Rev. 3 BFN-MD-N0039-880330, Engineering Evaluation to Qualify an Untested Seal Design - External Cable Seal, Rev. 0 BFN-MD-Q0100-89005, Evaluation of Internal Conduit Smoke and Gas Seal Design, Rev. 0 BFN-MD-QO1 00-980006, Evaluation of Penetration Seals, dated April 15, 1998 BFN-ND- Q0999-920115, Appendix R, Locations of Emergency Lighting, Rev. 3 BFN-ND- Q3999-930023, Unit 3, Appendix R Fire Suppression Damage Evaluation, Rev. 2 EDQ0999-940040, Appendix R Computerized Safe Shutdown Separation Analysis, Rev. 11 General Design Criteria No. BFN-50-747, Fire Protection of Safe Shutdown, Rev. 4 MD-Q0031-880249, 250 V Battery A, B, C, & D Ventilation Requirements, Sect. 3.5, Rev. 3 MD-Q0031-000007, Control Bay and Electric Board Room TMG Analysis, Rev. 2 TVAN Fire Protection License Condition Impact Evaluation (LCIE), Fire Protection Report, Rev. 20, dated October 10, 2002 Audits and Self-Assessments BFN-OPS-03-009, Self-Assessment, dated August 26,2003 Design Criteria and Standards BFN-50-0747, General Design Criteria for Fire Protection of Safe Shutdown, Rev. 4 BFN-50-0799, General Design Criteria for Fire and Pressure Seals, Rev. 4 BFN-50-7026, General Design Criteria for High Pressure Fire Protection System, Rev. 4 BFN-50-7308, General Design Criteria for Fire Alarm and Detection System, Rev. 1 Completed Surveillance Procedures and Test Records 0-SI-4.11 .A.1.a (3), Fire Detection Operability Test, Rev. 2, dated March 20, 2003 0-SI-4.11.G.1.(a), Visual Inspection of Fire Rated Barriers, Rev. 15, dated February 18, 2002 0-SI-4.11 .G.1.b(1), Visual Inspection of First Period Appendix R Fire Dampers, Rev. 7, dated October 10, 2001 0-SI-4.11 .G.2, Semiannual Fire Door Inspection, Rev. 20, dated May, 1, 2003 FP-2-247-INS003B, Emergency Lighting 18 Month Battery Discharge Test, Rev. 13, dated July ?0 2003 Fj-2-247-lNS004, Emergency Lighting Quarterly Functional Test, Rev. 19, dated September 8, 2003 Technical ManualsNendor Information Press-FD-1, Press Power Group, 3-Hr. UL Rated Internal Expansion Fire Damper, Rev. 1 Test Report CTP-1001 A, Southwest Test Research Institute, Three Hour Fire Qualification Test, 10" and 60 Depth Silicone RTV Foam Electrical and Mechanical Penetration Seals, dated July 25, 1980 ATTACHMENT

7 ATTACHMENT

8 Test Report CTP-1 024, Southwest Test Research Institute, Three Hour Fire Qualification Test for Electrical and Mechanical Penetration Seals, dated June 16,1982 Test Report CTP-1 040, Southwest Test Research Institute, Differential Pressure Test, Light Density Silicone Elastomer, dated December 14,1982 Applicable Codes and Standards NFPA 13, Standard for the Installation of Sprinkler Systems, 1976 Edition NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1974 Edition NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1987 Edition NFPA 72E, Standard on Automatic Fire Detectors, 1990 Edition NFPA 80, Standard on Fire Doors and Windows, 1975 Edition NUREG-1552, Supplement 1, Fire Barrier Penetration Seals in Nuclear Power Plants, dated January 1999 OSHA Standard 29 CFR 1910, Occupational Safety and Health Standards, Underwriters Laboratory, Fire Resistance Directory, January 1998 Other Documents 10 CFR 21-001, ELECTRAK Corporation, Software Error within TRAK2000 Cable Management and Appendix R Analysis System, dated March 7, 2003 BFN Simulator Exercise Guide OPLI73S149, Safe Shutdown Instructions - 2/3-SSI-9, Rev. 1 Factory Mutual Research Corporation (FM), Examination and Tests for Flamemastic 71A Cable Coating from Dyna-Therm Corporation, dated July 22, 1970 Fire Brigade Drill Data Sheets for period January 2001-May 2003 Fire Brigade Pre-plan No. CB3-617, Control Building Unit 3, Elevation 617', Fire Area 16, Rev. 2 Fire Brigade Pre-plan No. RX2-621, Reactor Building Unit 2, Elevation 621', Fire Area 9, Rev. 3 Fire Brigade Pre-plan No. RX2-593, Reactor Building Unit 2, Elevation 593', Fire Area 2, Rev. 3 Fire Brigade Pre-plan No. RX3-621, Reactor Building Unit 3, Elevation 621', Fire Area 13, Rev. 3 Fire Brigade Pre-plan No. RX3-639, Reactor Building Unit 3, Elevation 639', Fire Zone 3-4, Rev. 3 Fire Protection Weekly Inspection Reports for September 2003 Fire Reports and Investigations for May 2002 to September 2003 NRC Information Notice 2003-08, Potential Flooding through Unsealed Concrete Floor Cracks, dated June 25, 2003 Response to NRC Information Notice 1997-48, Inadequate or Inappropriate Interim Fire Protection Compensatory Measures (CMR-97-024), dated September 8,1997 U.S. Consumer Product Safety Commission, Invensys Building Systems Announce Recall of Siebe Actuators In Building Fire/Smoke Dampers, dated October 2, 2002 License Basis Documents BTP Chemical and Material Engineering Branch CMEB 9.5-1 Letter, dated July 1981 Fire Protection Report Volume 1, Fire Protection Plan, Rev. 23 Fire Protection Report Volume 1, Section 2, Fire Hazards Analysis, Rev. 23 Fire Protection Report Volume 1, Section 3, Safe Shutdown Analysis, Rev. 23 Fire Protection Report Volume 2, Section I-A, Fire Reports and Investigations, Rev. 0 Fire Protection Report Volume 2, Section l-D, Smoking Restrictions, Rev. 0 ATTACHMENT

9 Fire Protection Report Volume 2, Section l-G, Storage and Labeling of Hazardous Chemicals, Flammable or Combustible Liquids, and Compressed Gas Cylinders, Rev. 0 Fire Protection Report Volume 2, Section 1II,Fire Brigade Training and Fire Drill Evaluations/

Critiques, Rev. 2 NRC Safety Evaluation Report dated March 31, 1993 NRC Safety Evaluation Report dated November 2, 1995 Summary of Deviations from NFPA Code for BFN, dated August 3, 1988 Updated Final Safety Analysis Report, Section 10.18, Plant Communications System, Rev. 19 Updated Final Safety Analysis Report, Section 10.19, Lighting System, Rev. 18 PERs Reviewed 03-001375-000, Diesel Driven Fire Pump Temperature Records 03-002935-000, Fire Pump Failed Capacity Test 03-008165-000, Evaluate Heat Collectors Over Sprinklers Per IN 2002-24 03-009529-000, Asiatic Clams Found in Yard Fire Protection System 03-013828-000, Procedure MMDP-1 Does Not Consider Impacts on Fire Protection Administrative Controls 03-013882-000, NRC Letter SECY-03-100 Was Recently Issued on Rulemaking for Manual Actions Used for 10 CFR 50, Appendix R, Section III.G.2 Compliance.

PERs and Work Orders Generated Durinq this Inspection 03-000461-000, TVA calculation Issued referencing another non-approved calculation.

03-016883-000, BFN-0-PMP-026-0003 Pump Packing Leak, Fire Pump C 03-017102-000, BFN-0-ISV-026-0565 Valve Packing Leak, Fire Pump A Discharge Shutoff Valve 0$-017292-000, Smoke detector 0-SDE-26-87JW Is installed In the Incorrect location from that fhown on location plan 0-47W600-268 (Fire Area 13, location plan) 03-017479-000, Procedure changes needed for 0-AOI-26-1, 2-AOI-1 00-1, and 2/3-SSI-001 03-018587-000, Channel Diesel Fire Pump fill valve was not locked in the open position 03-018593-000, Generic review of SQN PER 03-011569-0 on NRC concerns regarding fire protection compensatory measures 03-018973-000, Administrative discrepancies in the Fire Protection Report 03-019088-000, Typographical errors Identified In 2/3-SSI-3-4 03-019089-000, 2/3-SSI-1 3, Attachment 6, does not specify that a ladder may be required to operate valve 3-BYU-84-686 03-019164-000, Tamper-proof covers for the Appendix R switches that operate the RHR injection valves could not be operated without the use of a tool 4i-019210-000, Evaluate 0-AOI-26-1 for enhancements with regard to using auxiliary equipment for smoke removal 03-019211-000, PER to track the evaluation of the associated circuit issue at BFN as identified at Hatch involving spurious SRV opening due to fire effects on pressure transmitters 03-019212-000, PER to track a URI at BFN with respect to multiple spurious actuations resulting from a fire. BFN does not assume that any one spurious actuation or signal can adversely affect multiple valves in series 03-019227-000, Definitions in various calculations and documents are not consistent ATTACHMENT

10 ATTACHMENT

11 03-019229-000, Hydrogen build up Inthe shutdown battery rooms C and D as a result of loss of exhaust capability was not adequately evaluated and documented 03-019230-000, Fire-induced circuit faults associated with valve 2-FCV-74-106 (RHR pump drain valve) and its Impact of RHR Pump 2D start capability was not adequately documented ATTACHMENT

12 LIST OF ACRONYMS AC alternating current ADAMS Agency-Wide Documents Access and Management System AOI abnormal operating instruction BFN Browns Ferry Nuclear BTP Branch Technical Position CAP corrective action program CFR Code of Federal Regulations CMEB Chemical and Material Engineering Branch DC direct current ELU emergency lighting unit FHA Fire Hazards Analysis FPP Fire Protection Program FPR Fire Protection Report GL gGeneric ILetter HVAC heating, ventilation, and air conditioning kV kilovolt MOV motor operated valve NCV non-cited violation NFPA National Fire Protection Association NRC U.S. Nuclear Regulatory Commission OSHA Occupational Safety and Health Administration PARS Publicly Available Records Systems PER problem evaluation report RHR residual heat removal RRP reactor recirculation pump SER safety evaluation report SPP Standard Programs and Processes SRV safety relief valve SSAR Safe Shutdown Analysis Report SSD safe shutdown SSI safe shutdown instruction TS Technical Specification(s)

TVA Tennessee Valley Authority TVAN TVA Nuclear UFSAR Updated Final Safety Analysis Report URI unresolved Item V volt ATTACHMENT

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my review comments on BF IR 03-07 Creation Date: 12/1/03 2:30PM From: Charles R. Ogle Created By: CRO@nrc.gov Recipients Action Date & Time nrc.gov ATlPO.ATLDO Delivered 12/01/03 02:30PM DCP (Charlie Payne) Opened 12/01/03 02:30PM Post Office Delivered Route ATL_PO.ATL_DO 12/01/03 02:30PM nrc.gov Files Size Date & Time comparison_bf_Rev3_to_final.wpd 202764 12/01/03 02:28PM MESSAGE 1598 12/01/03 02:30PM Options Auto Delete: No Expiration Date: None Notify Recipients: Yes Priority: Standard Reply Requested: No Return Notification: None Concealed

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