Text

IR 05000260-03-007 and IR 05000296-03-007 on 09/08/03 - 09/12/03 and 9/29/03 - 10/03/03 for Browns Ferry, Units 2 and 3; Triennial Fire Protection
	ML050540213
Person / Time
Site:	Browns Ferry
Issue date:	11/17/2003
From:	Ogle C; Division of Reactor Safety II
To:	Scalice J; Tennessee Valley Authority
References
	FOIA/PA-2004-0277 IR-03-007
	Download: ML050540213 (35)
	v • d • e

Olr--I UNITED STATES

o - NUCLEAR REGULATORY COMMISSION

REGION II

SAM NUNN ATLANTA FEDERAL CENTER

$ 961 FORSYTH STREET SW SUITE 23T85

ATLANTA, GEORGIA 30303-8931

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 certain non safety control

circuit cables and instead using unapproved,procedural guidance directing a local manual

operator actions in the Unit 3 480 Wvolt Reactor Motor Operated Valve Board Room 3ARoom

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 concern. 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 in this 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 htto://www.nrc.gov/

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

Sincerely,

/RAI

Charles R. Ogle, Chief

Engineering Branch 1

Division of Reactor Safety

Docket Nos.: 50-260, 50-296

License Nos.: DPR-52, DPR-68

Eneclosm e: (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 wencl cnt'd See page 3)

(ee Wen-elent'd)

State Health Officer

Alabama Dept. of Public Health

RSA Tower - Administration

Suite 1552

P. 0. Box 303017

Montgomery, AL 36130-3017

Chairman

Limestone County Commission

310 West Washington Street

Athens, AL 35611

Distribution w/encl:

K. Jabbour, NRR

L. Slack, RlI EICS

I BIDSnRDSNnRDlrMLIrBRIDSNRRDIPMLIPB

PUBLIC

TVA 4

OFFICE RII:DRS RII:DRS RII:DRS CONTRACTOR RII:DRP

SIGNATURE RA

RARA RA

NAME SWALKER GWISEMAN CPAYNE KSULLIVAN SCAHILL

DATE 11/10/2003 11/1712003 11/17/2003 10/28/2003 - 11/17/2003

E-MAIL COPY? YES YES NO YES NO YES NO YES NO YES NO YES NO

PUBLIC DOCUMENT YES NO KI _________ _ As

IIII _ _ _ ___.______

UPFILIAL NtLUUtlU tuAY L)GUMU NTI I NAME: R:=omparison-oT-tevJ toinai.wpa

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.: 50-260, 50-296

License Nos.: DPR-52, DPR-68

Report No.: 05000260/2003007 and 05000296/2003007

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 Intern

Approved by: Charles R. Ogle, Chief

Engineering Branch 1

Division of Reactor Safety

Enclosure

SUMMARY OF FINDINGS

IR 05000260/2003-007, 05000296/2003-007; 9/8 - 12/20039/12/2003 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-1 649, "Reactor Oversight Process," Revision 3,

dated July 2000.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: 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 licensee 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 1-3

(Unit v 400 V fleaeteo Motor Operated Valve ocard floom 3A). in lieu of

protecting these cables, the licensee used manual operator actions to locally trip

the Unit 3 RRPs but failed to obtain prior NRC approval. Additionally, during a

severe fire in Fire Area 13, these manual actions would be performed 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

actions are performed.

This finding is unresolved pending completion of a significance determination.

This finding is greater than minor because it adversely impacted the capability of

systems, structures and components necessary to achieve and maintain the

plant in a saf e shutdown, Wndition during a severe fire and affeteudis 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)

Green. A Severity Level IV non-cited violation (NCV) of 10 CFR 50.48(a) and

the Unit 2 and 3 Operating License Conditions was identified for the licensee

making a change to the approved fire protection program (FPP) which removed

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 1R05.1 1)

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]fFPP; 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 Required 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 in each 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.

Verify that local manual operator actions are consistent with the plant's fire

protection licensing basis.

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-1 00-1, 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 Reactor Motor Operated Valve (RMOV) Board

Room 3A

EPIP-17, Fire Emergency Procedure

b. Findings

Introduction: A finding was identified in that the licensee failed to protect the men safety

control circuit cables forSafe 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 3 480 V FReactor Motor Operated Valve Board Roorn 3A).

In lieu of protecting these cables, the licensee used manual operator actions to locally

trip the Unit 3 RnPs but failed to obtain prior NfC approval. Additionally, during a

severe fire in Fire Area 13, these manual actions would be performed in Fire Area 13

axd. 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 aeti&-afreaction 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 in the main

control room. However, in certain fire areas, RRP control power eircuitscables 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-1 3 directed the operator to enter Fire Area

13, go to the 250 V DC Reactor Motor Operated Valve (RMOV) Board 3A (located in

Fire Area 13), and place the 4160 V RPT Board 3-l1 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 in 2/3-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 concluded that-the

nnr control power circuits were associated non safety circuits as defined in 10 c R50o,

Appendix R, Section 11I.G.2 . A review of the Browns Ferry SERs found that these

manual operator actions had not been approved by the NfC for use in lieu of protecting

the eabfes.

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-03-1 00. 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 capability of systems, structures and

components necessary to achieve and maintainSSI procedure quality for achieving and

maintaining the plant in a 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 NlC approved FPP fer Drowns Ferry Units 2 and 3 commits to 10

OFR 50, Appendix R, Section 111.G. Section 11I.G.2 states, in part, '"here cables and

equipment including associated non safety circuits that could prevent operation or cause

maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of

systems necessary to achieve and maintain hot shutdown conditions are located within

the same fire area outside of primary containment, one of the following means of

ensuiring 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 no intervening combustibles or fire hazards plus fire

4

detection and an automatic fire suppression system; or (3) physical protection by a one

hour rated fire barrier plus fire detection and an automatic fire suppression system."

Manual operator aetion to respod to maleperations is et listd a aaeptable

method for satisfying this requirement.

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 licensee failed

to protect the non safety control circuit cables for the Unit 3 RRPs whose maloperation

during a severe fire could pravent the redundant train of safe shutduwn systems from

successfully achieving and maintaining hot shutdown conditions. In lieu of providing

adequate physical protection, ,thelicense use manual2/3-SSI-13 was not properly

established, in that certain operator actions outside the main control room without

obtaining prior Nzic approval. In addition, these manual operator actionsprescribed 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 Fire Damage And Unapprovedinadequate Unit 3 Fire Procedure

Directs Local Manual Operator Actions Be Performed In Location of Fire.

.02 Fire Protection of Safe Shutdown Capability

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 I-D, Smoking Restrictions

TVAN Standard Programs and Processes Procedure SPP-1 0.10, Control of

Transient Combustibles

TVAN Standard Programs and Processes Procedure SPP-1 0.11, Control of

Ignition Sources (Hot Work)

Electrical Preventive Instruction EPI-0-000-MCC001, 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 in the

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 in support 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 in a 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 in the 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 in this

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 in support 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 in plant

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 addtion, the inspector

reviewed a sampleCf theI IVAG system for the selected f ire area-sa.

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 III.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 Alternative Shutdown Capability

a. Inspection 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 in the intervening period. Thus,

alternative shutdown capability was not reviewed during this inspection.

b. Findings

No findings of significance were identified.

.05 OQerational 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.

l .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

eoisistent-withper 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 Emergency Lighting

a. Inspection Scope

The inspectors reviewed the design, operation, and manufacturer's 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 IIU.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

No findings of significance were identified.

I

9

.08 Cold Shutdown Repairs

.a Inspection Scope

The licensee had identified no need forlicensee'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) O-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 in which they were installed. The inspectors

reviewed the installation instructions for fire doors, the design details for mechanical and

electrical penetrations, the penetration seal database, gGeneric 1Letter (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 in support 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 Equipment

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 in the 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 Scone

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 1Il.G.2 and Ill.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-1 0.9 (Control of Fire Protection Impairments,) 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;

anld

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. In this

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 and/or Spr inkle, Systems, Seeton 9.3A .11 3.,6%.-Systcm cand -Seticn

9.3.11.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, Frn Section 7.5

was revised to delete the requirement for a fire watch for areas in which the fire

suppression equipment was inoperable.

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 III.G.3, without establishing compensatory fire watches

being established in the affected plant areas ifas long as fire detection systems were

funetioningfunctional. The change adversely affected the ability to achieve and maintain

safe shutdown in the event of a fire, in that, the licensee went from full compliance with

the fire protection safe shutdown system separation and suppression criteria to less

than full compliance without implementing eompentetytemporary 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.(14) and Browns Ferry Unit 3 Operating License Condition 2.C.(7) state, in part, 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 under the

GDP because it impacted 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 in accordance 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

l 17, 2003, to provide an update on changes to the preliminary inspection findings. The

I 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 Licensing 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 To Protcet Unit 3 Reactor Recirculation Pump

ConrolC~ruitry Firom Fire Damnage And

UmapprovedInadequate Unit 3 Fire Procedure Directs

Local Manual Operator Actions Be Performed In Location

of Fire (Section 1R05.01)

ODened and Closed

05000260,296/2003007-02 NCV Changes Made to the Fire Protection Program Regarding

Compensatory Fire Watch Implementation Without NRC

Approval (Section 1R05.1 1)

Discussed

None

ATTACHMENT

2

LIST OF COMPONENTS INSPECTED

I Section 1R05.02: Fire Protection of Safe Shutdown Capability

I

TexC Moved Idere: 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 Of Moved Text

Section 1R05.03: Post-Fire Safe Shutdown Capability

Component Identification Description

0-PMP-026-0001 'A' Electric Fire Pump

0-PMP-026-0002 'B' Electric Fire Pump

0-PMP-026-0003 'C' Electric Fire Pump

O-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-3-204M Main Steam Pressure Switch Div. I

ATTACHMENT

3

PS-3-204CB Main Steam Pressure Switch Div. 1I

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/Zone/Room 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-2010, 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 From l lere: 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-4.1 1.E.1.b(1), Fire Hose Station Operability/Flow Test, Rev. 3

1-ARP-9-20-A, Alarm Response Procedure, Rev. 14

1-ARP-25-165, Alarm Response Procedure, Rev. 13

2-AOl-1 00-1, 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 0.10, Control of Transient Combustibles, Rev. 2

TVAN SPP-10.1 1, Control of Ignition Sources (Hot Work), Rev.1 B1

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, & AirFire 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-1 28, Rev. 1, Sh. 2ES-1 46, 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 1-1, Flow Diagram Residual Heat Removal System, Rev. 58

2-47E850, 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-1 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

3-47E61 1-1-1, 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-11F, 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-QO100-89005, Evaluation of Internal Conduit Smoke and Gas Seal Design, Rev. 0

BFN-MD-Q0100-980006, 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 20, 2003

FP-2-247-lNS004, Emergency Lighting Quarterly Functional Test, Rev. 19, dated

September 8, 2003

Technical Manualsl/endor Information

Press-FD-1, Press Power Group, 3-Hr. UL Rated Internal Expansion Fire Damper, Rev. 1

Test Report CTP-1 001A, Southwest Test Research Institute, Three Hour Fire Qualification

Test, 10" and 6" 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

ML050540213

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