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NUREG-O061

SAFETY EVALUATION FOR OPERATIONS

OF BROWNS FERRY, UNITS I AND 2,

FOLLOWING THE MARCH 22, 1975, FIRE

DATE PUBLISHED: MARCH 1976

Division of Operating Reactors

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

TABLE OF CONTENTS

PAGE

1.0 Introduction ........................................... 1-1

2.0 Results of Investigation of Damage ..................... 2-1

2.1 Description of Fire Affected Area ................. 2-1

2.2 Circuits Damaged .................................. 2-2

2.3 Electrical Equipment Other Than Cable ............. 2-3

2.4 Evaluation of the Integrity of Piping, Structural

and Mechanical Components for Continued Operation 2-5

2.4.1 Piping ..................................... 2-5

2.4.2 Mechanical and Structural .................. 2-7

2.4.3 Summary of Inspection Results .............. 2-10

2.5 Cable Replacement and Repair ...................... 2-11

3.0 Approach for BFNP Fire Protection ...................... 3-1

4.0 Plant Fire Protection Reanalysis ....................... 4-1

5.0 Fire Protection System Impairment Analysis ............. 5-1

6.0 Administrative Changes ................................. 6-1

6.1 Training (Fire Related) ........................... 6-1

6.2 Procedures (Fire Related) ......................... 6-3

6.3 Organization ...................................... 6-7

6.4 Quality Assurance ................................. 6-9

7.0 Design Modifications ................................... 7-1

7.1 Electrical Changes ................................. 7-1

7.2 Use of Flamemastic 71A (Fire retardant coating) .... 7-3

7.3 Cable Spreading Room Ventilation .................. 7-5

i
TABLE OF CONTENTS (cont'd)

PAGE

7.4 Fire Stop Designs .................................. 7-6

7.5 Design Changes Related to Fire Detection and

Extinguishment that are to be Completed Prior

to Startup ......................................... 7-8

7.5.1 Fixed Water Systems .......................... 7-8

7.5.2 Heat Detection System ....................... 7-10

7.5.3 Smoke Detection System ...................... 7-10

7.5.4 Self-Contained Breathing Equipment .......... 7-11

7.5.5 Communication System ........................ 7-12

7.5.6 Emergency Lighting .......................... 7-13

7.5.7 Ventilation Systems ......................... 7-14

7.6 Design Changes Related to Fire Detection and

Extinguishment Intended to be Completed at the

First Refueling Outage of Either Unit .............. 7-15

7.6.1 Automatic Deluge Systems .................... 7-15

7.6.2 Automatic Pre-action Sprinkler Systems ...... 7-16

7.6.3 Manual Dry Pipe Sprinkler System ............ 7-17

7.6.4 Automatic CO2 System in Cable Spreading Rooms 7-17

8.0 Technical Specifications ................................ 8-1

9.0 Retest Program .......................................... 9-1

9.1 Fire Protection Systems ............................ 9-1

9.2 Program for Component and System Retests ........... 9-2

10.0 Conclusions ............................................. 10-1

ii
TABLE OF CONTENTS (cont'd)

ATTACHMENTS

ATTACHMENT 1 CHRONOLOGY

ATTACHMENT 2 ROLE OF CONSULTANTS

ATTACHMENT 3 NRC COMMENTS ON NELPIA

RECOMMENDATIONS

iii

1-1

1.0 Introduction

On March 22, 1975, a fire at the Browns Ferry Nuclear Plant caused

a shutdown of Units 1 and 2. The facility subsequent to the shutdown was found to have incurred substantial damage to power, control,

and instrumentation wiring. All three units are presently in the

shutdown condition with the fuel removed from the vessels for Units

1 and 2; the Unit 3 reactor is still under construction with operation

for that unit scheduled for early 1976. An overall program has been

developed by the licensee delineating the necessary activities required

to restore damaged portions of the facility to a safe level so that

operation of Units I and 2 can be resumed. This program has been

described in "Plan for Evaluation, Repair and Return to Service

of Browns Ferry Units 1 and 2 (March 22, 1975 Fire)" dated April 13,

1975, and revisions thereto up to and including Revision 37 (the Plan).

As a result of the fire an NRC review program was developed consisting

of three major and parallel elements. The first element was the

investigation conducted by the Office of Inspection and Enforcement

(OI&E) of events leading to the fire, fire fighting efforts, sequence

of operational events and problems experienced with the nuclear steam

supply system, interaction between units, and the response of TVA, and

State and Local authorities. This phase of the NRC program has been

completed and the results are given in the OI&E Investigation Report

of the March 22, 1975 Fire at the Browns Ferry Nuclear Station, dated

July 25, 1975.

The second element of the program, being performed by the Office of

Nuclear Reactor Regulation has as its objectives (1) to assure that

a safe plant configuration was attained and is being maintained

subsequent to the fire, (2) to assure safety during removal of fuel

from Units 1 and 2, (3) to assure plant safety during fire damage

removal and restoration, and (4) to determine that the design changes

that are required for restoration of these plants to operational

status are acceptable. (See Attachment 1 for Chronology)

The third element of the NRC program consists of the review being

performed by the Special Browns Ferry Review Group established on

March 26, 1975. Efforts of this review group will be to prepare for

NRC, recommendations to change, as required, NRC policies, procedures,

and technical requirements in regard to fire protection at nuclear

'power plants. The Review Group's report is anticipated about March 1,

1976.
1-2

The NRR reviewers of Browns Ferry Plant restoration and modifications

have had the benefit of informal discussions with members of the

Special Review Group (SRG). In this way, the NRR reviewers have been

aware of the recommendations to be made by the SRG. These recommendations were considered and appropriately implemented in Browns Ferry.

In some cases, the SRG recommended that further generic consideration

be given to certain aspects of fire protection or recommended that

consideration be given to changing NRC requirements or guidelines. In

the areas that require additional future consideration, the Browns

Ferry modifications will be reevaluated when any new position has been

developed.

The SRG also had the benefit of the ongoing TVA and NRC evaluations of

the Browns Ferry fire and the proposed restoration and corrective

modifications. Copies of the Plan and all revisions thereto were

supplied to the SRG, as well as NRR correspondence, meeting summaries,

and safety evaluations for NRR licensing actions during the course of

restoration.

A comparison of the Special Review Group's recommendations with the

modified Browns Ferry Plant will be presented in a supplement to this

report.

Thus far there have been three significant licensing actions taken

to meet the objectives of the review performed by the Office of

Nuclear Reactor Regulation. First, the plant Technical Specifications

were changed to Temporary Technical Specifications designed to assure

that a safe configuration was maintained following the fire, while

further corrective action plans were under development. These

Temporary Technical Specifications and the associated safety

evaluation were issued with Amendment 9 to License DPR-33 (Unit 1)

and Amendment No. 6 to License DPR-52 (Unit 2) on May 9, 1975.

The Technical Specifications were again changed to provide for the

removal of fuel from both Unit 1 and 2 reactor vessels and placement

into the respective storage pools. These changes and their associated

safety evaluation were issued with Amendment 10 to License DPR-33

and Amendment 7 to License DPR-52 on June 13, 1975. These changes

also assure that the necessary protective equipment to assure

protection of- the fuel in the storage pools from damage would be

available with adequate redundancy and would be protected against

adverse interaction with the fire affected wiring and cables, while

such fire damage was removed and when the damaged wiring and cabling

was replaced.
1-3

On September 2, 1975, NRC issued Amendment 14 to License DPR-33 and

Amendment 11 to License DPR-52 that authorized restoration and

modification activities described in the Plan up to and including

Revision 20. The safety evaluation accompanying those amendments

describes the work that was authorized and presents our conclusions

as to its acceptability. Part of our conclusion at that time was

that additional measures would be required to enable us to find that

the plant was acceptable for return to operation. By letter dated

August 29, 1975, TVA agreed to provide further improvements in fire

protection. The revisions to the Plan from Revisions 22 to 37 included

provisions of the implementation of the commitments made in the

August 29, 1975 letter.

This Safety Evaluation Report presents the results of our evaluation

regarding the acceptability of the restoration and modifications

at the Browns Ferry Nuclear Plant Units 1 and 2 to establish that the

facility may be operated in the restored and modified condition without

endangering the health and safety of the public. Although some of these

activities were the subject of the September 2, 1975 evaluation, a

complete evaluation will be presented herein. This will present a

total picture of the substantial enhancement of the capability of

the facility that is now provided to prevent and withstand a fire.

In our evaluation, we made use of our own consultants and those of the

licensee. (See Attachment 2)

The TVA modification program related to fire detection and extinguishment

is being conducted in two stages. The first stage covers those changes

to be accomplished prior to restart, which is currently scheduled by TVA

for the end of March 1976. Section 7.5 of this report discusses these

changes and provides the staff evaluation for approval. Section 7.6

discusses the second stage modifications which are intended to be completed at

the first refueling outage of either unit. It provides the staff evaluation

for approval of the design criteria.

A letter from the Office of Inspection and Enforcement will provide

confirmation that all of the work described in the Plan as being

required prior to restart has been completed before the authorization

for return to operation will be issued.

2-1

2.0 Results of Investigation of Damage

2.1 Description of Fire Affected Area

Substantial detail concerning the cause of the fire and the extent

of damage to structures, systems and components is provided in

the previously noted OI&E report. The following brief summary is

provided as background for the discussion of the repair and replacement activities described in this Safety Evaluation.

After ignition, the fire propagated through the penetration in

the wall between the cable spreading room and the Unit 1 reactor

building. As a result of the pressure differential which is

maintained between the reactor building and the cable spreading

room, and because an installed carbon dioxide (C0 2 ) fire

extinguishing system was used in the cable spreading room, only

a small amount of burning occurred in the cable spreading room.

Damage to the cables in this area was limited to a few feet

immediately adjacent to the penetration where the fire started.

The major damage occurred in the reactor building outside the

cable spreading room, in an area roughly 40 feet by 20 feet, where

a very high concentration of electrical cables exists. There was

very little other equipment in the fire affected area, with the

only direct damage other than cables being the melting of a soldered

joint on an air line.

In order to provide a basis for assessing damage within the fire

affected area TVA developed a program for determining temperature

zones. Six zones were established on the basis of physical evidence

ranging from buff colored concrete (1500 0 F) to paint softened but

not blistered (3000 F). These zones were utilized by other study

teams to determine the temperatures to which various pieces of equipment may have been subjected.

The investigation of damage was conducted throughout the reactor

building in both Units because the spread of soot and products of

combustion extended beyond the immediate fire affected area.
2-2

2.2 Circuits Damaged

The electrical cables, after insulation had been burned off, shorted

together, and grounded to their supporting trays or to the conduits,

resulting in the loss of control power for much of the installed

equipment such as valves, pumps, blowers and the like.

Exhibit Cl of the OI&E report is the TVA report of physical damage

to the electrical cables and raceways. It identifies 1611 cables

that were damaged or assumed damaged. Exhibit C3 of the OI&E report

lists the Unit 1 systems lost during the fire resulting from cable

damage.

The most significant aspect of the damage was that it resulted in

the loss of redundant safety equipment. However, adequate cooling

to remove decay heat was always available. Our assessment of plant

safety during and after the fire was included with Mr. Benard Rusche's

testimony at the September 16, 1975 Hearings before the Joint Committee

on Atomic Energy concerning the Browns Ferry fire,

As discussed in Section 3 of this SER, although the units were safely

shut down and cooled, the staff has required considerable change

in administrative actions, isolation* of redundant safety circuits,

and fire detection and extinguishment capabilities. These changes

minimize the possibility of loss of more than one division of cooling

equipment.

The term isolation of redundant circuits is used in this report to

mean those design features, such as barriers or separation, employed

to prevent a fire in one division from affecting the other division.
2-3

2.3 Electrical Equipment Other Than Cable

The Browns Ferry damaged electrical equipment identification program

encompasses a three pronged approach. The investigation considers

equipment exposed to high temperatures, equipment contaminated by

smoke and soot, and equipment exposed to abnormal electrical power

conditions. An elevated temperature zone has been established by visual inspection of the areas around and adjacent to the fire. All

electrical equipment within the established zone has been identified

and thoroughly checked for any degradation. Either during or following

cleanup of the identified equipment, a visual inspection has been

made for signs of distress or overheating damage sustained by

insulating components. If necessary, components were disassembled

to verify observations. Non-destructive electrical tests appropriate for the type of equipment involved will have been made and the results

will have been evaluated prior to startup. Preoperational tests

will then be performed to verify the correct functional performance

of the equipment, controls, or circuits involved. (See Section 9.2).

A larger zone than the high temperature zone has been established

for the purpose of removing soot and smoke residues from electrical

equipment. Equipment in this zone beyond the boundaries of the

high temperature zone also underwent physical inspection during

cleanup. The equipment will undergo a similar verification program

as outlined above for equipment in the high temperature zone prior

to operation.

The third area of investigation is abnormal electric power conditions.

Due to the extensive cable damage in the cable trays and the fact

that most cables were energized, the possibility exists that

excessive transient voltages and current, short circuits, and

ground faults occurred. The Plan states that all power circuits

and most control circuits are protected by relays, fuses or circuit

breakers which operate for overcurrent or fault conditions and

thus protect the electrical equipment. While this is true, absolute

credit for overcurrent devices.cannot be given under these

circumstances. We have taken the position that with the possibility

of hot shorting within the trays, proper circuit protective action

may not have been effective on some circuits. TVA committed to perform

diagnostic electrical tests on all electrical equipment that may

have been exposed to abnormal electrical power conditions. This

is a result of our position that visual inspection alone was

insufficient to ensure all degraded electrical equipment would

be identified. TVA states that this requirement will be satisfied

by performing the inspections and tests specified in the cleaning

procedure*or by performing satisfactory functional tests after

cable restoration.

*Part V, Section B.2 of the Plan.
2-4

Any equipment found damaged, to have failed, or to be unsuitable

for continued service shall be repaired if repair is technically

feasible, not extensive, and economically justifiable. Equipment

not fitting into the above categories will be replaced. When

repairs or replacements are completed, surveillance tests or preoperational retests will be performed to demonstrate satisfactory

performance. To date no fire related damage to electrical equipment

other than cable has been found.

We find this program gives reasonable assurance that all damaged

electrical equipment will be identified and properly restored and

is therefore acceptable.
2-5

2.4 Evaluation of the Integrity of Piping, Structural and Mechanical

Components for Continued Operation

We have considered potential adverse effects to the metallic materials

of safety related mechanical domponents from heat and combustion

product release from the fire. Effects identified as requiring

detailed evaluation were (1) stress corrosion due to chloride contamination from combustion products, (2) liquid metal embrittlement

from liquefaction of low melting point metals, and (3) elevated

temperature exposure effects on materials.

The burning of the polyvinyl chloride electrical insulation produced

an estimated 1400 pounds of chloride in the form of hydrochloric acid

droplets mixed into a soot-like residue. Chloride containing residue

was deposited throughout the Unit 1 reactor building and in limited

areas in the Unit 2 reactor building. Chlorides are corrosive to

most materials but the principal concern is the possibility of their

initiating stress corrosion cracking in austenitic stainless steels

or high strength ferritic steels. Chemical analysis of the residue

indicated that it was 20% chloride with zinc, lead, tin and copper

each present in concentrations of less than one percent.

The licensee identified all components susceptible to possible stress

corrosion damage and implemented a program of cleaning, cleanliness

verification, inspection, evaluation and surveillance. The combustion

product residue was removed by dry cleaning followed by wet detergent

cleaning. The procedures were repeated until measured chloride levels

were below 0.08 mg per dm2 . This is recognized as an acceptable

level for chloride in contact with stainless steel and is incorporated

as the acceptance level in ERDA/RDT standard Number FS-IT, "Cleaning

and Cleanliness Requirements for Nuclear Components." All components

(except electrical) have been cleaned to this standard.

After cleanliness was verified, precautions were taken to prevent

recontamination. Locations where chloride contamination could have

been trapped were identified. For example, porous thermal insulation

was removed and replaced with new insulation. Metal covered non-porous

insulation has little potential for acting as a trap for contamination

but it will have been tested to assure that chlorides have not penetrated.

2.4.1 Piping

Insulated and uninsulated piping were both used within the zone of

influence of the fire. Since all uninsulated piping within the heat

affected zone contained water during the fire their temperature

increases were limited. All piping was evaluated on the basis of

temperature effects and chemical corrosion effects before acceptance

for continued service. Replacement with new piping was used as

an alternative to the evaluation procedure.
2-6

The possibility of austenitic stainless steel piping being heated

into the sensitization range (800 to 12000F), which would increase

its susceptibility to stress corrosion, was investigated. It was

determined that no stainless steel pipe was exposed to these

temperatures. In a few instances, carbon steel piping (4 inch diameter

maximum) was replaced because of heat effects. Any aluminum piping

determined to have been exposed to a temperature in excess of 340°F

was replaced. One run of aluminum pipe was replaced because of

pitting and corrosion. Penetrant inspection was performed on the

carbon steel and aluminum tubing on a sampling basis with no

significant problem areas detected. All small diameter copper

instrument lines in the region of the fire were found to be discolored

and were replaced without further evaluation.

Stress corrosion cracking related to the chloride contamination from

combustion products could occur immediately or in the future due to

residual effects. Each stainless steel safety-related component or

pipe that was exposed to possible chloride contamination was subject

to liquid penetrant examination and will be monitored by a surveillance

program. The penetrant inspection included 100% inspection of the

core spray line, control rod drive hydraulic return line, reactor

water cleanup line, reactor water level monitor line, HPCI and RCIC

instrumentation lines and each of the attached valve bodies. The

other stainless steel components were examined on a sampling basis

as described in Part V of the Plan, Section B.1, App. B to

Mechanical Maintenance Instruction 46, which we found acceptable.

No indications of damage were found and these items are now under

a surveillance program.

*To completely evaluate the effects of the combustion product residue,

11 metallurgical samples were removed from Unit 1 stainless steel

components. One sample from a high strength steel core spray hanger

spring was also taken. The examination included a boat sample

containing the weld and heat affected zone from the core spray

elbow. Samples were taken also from high residue areas outside

of the fire zone. Surface replicas are included in the program

because this is a non-destructive technique that would provide

greater sampling.

In addition to conventional metallography, scanning electron microscopy and qualitative spectroscopy were used for the evaluation.

Only one minor indication that could be attributed to the initial

stages of stress-assisted corrosion was found. This was slight

intergranular attack on a small diameter sampling line located

on the fire level. Additional metallographic evaluation will be

conducted on thin wall instrument tubes in the fire area with

results submitted for review to NRC prior to startup to confirm

that there is no damage.

The qualitative spectrograph is extremely sensitive to small

concentrations of elements. It detected trace levels of chlorine

on three samples. None of the embrittling elements such as zinc,

lead or tin were detected by the spectrograph. These elements were

found in the combustion product residue but to cause embrittlement,
2-7

they would have to be in contact with the steel while in the liquid

state. We conclude that liquid metal embrittlement is not a factor

and that the trace indications of chlorides are not significant.

While we conclude that the cleaning procedures and subsequent inspections

were adequate, there is no assurance that corrosion assisted effects

will not appear at some later time. The licensee has committed to a

surveillance program for the first two years. Quarterly, after the

unit is returned to power, a general visual examination of the piping

systems for evidence of leakage, cracking and other distress will be

conducted by qualified personnel. At the first refueling, liquid

penetrant inspection will be conducted in the same areas as in the

original inspections discussed above. In addition, surface replication

will be conducted in the same areas as the original metallographic

examinations; e.g., areas with heavy deposits of residue. Any adverse

findings will increase the scope of the examination.

We are in basic agreement with the licensee's surveillance program

approach, however, we are in the process of evaluating the need for

an increased scope and longer period of surveillance. The details

of the full program will be finalized prior to the return to operation

and will be included in a supplement to this report.

We have had the benefit of the comments and recommendations of the

licensee's two consultants (Part V of Plan) in the area of fire

effects on materials. We are satisfied that the licensee has been

responsive to these comments and recommendations. (See Attachment 2)

We have reviewed and evaluated the program that the licensee has

implemented to demonstrate the absence of detrimental effects on the

metallic materials of construction of safety related mechanical

components. We conclude that the licensee has conducted an adequate

program of investigation of problem areas, component cleaning, cleanliness verification, evaluation and testing.

2.4.2 Mechanical and Structural

Safety related heating ventilating and air conditioning ducts were

evaluated on the basis of exposure to temperature of 5OOOF or greater

as gaged by the established temperature zones and visible distortion

of the ducts. All ducts showing visual distortions have been replaced

or reworked to original specifications. All duct work influenced

by the fire has been cleaned regardless of temperature exposure.
2-8

Evaluation of cable trays, cable tray supports and fixed members of

pipe supports was based on exposure temperatures and stress analysis

considering material strengths reduced by any possible annealing affects

of higher temperature exposure. Any structural steel found to have

been immersed in a temperature of 1000OF or higher was replaced without

further evaluation.

In general all cable trays in direct and near contact with cables

actually consumed by the fire were replaced without further evaluation.

The larger cable tray supports were re-evaluated based on the actual

load to be sustained. The acceptable stress levels for these analyses

were set at either 75% of the original design allowable stress or

based on the measurement of actual physical properties from a sample

of the material exposed to the actual fire conditions.

The requirements specified above for acceptance of cable tray supports

were also employed for acceptance of the fixed members of pipe

supports. Additional requirements have been established for acceptance

of the variable elements of pipe supports. In the absence of variations in setting from the effects of the fire and the absence of actual

fire damage the support setting was corrected to give a properly

aligned piping system. Setting changes as a result of the fire with

no visible evidence of heat damage required evaluation of spring

characteristics to assure the necessary range of support as required

by the system design. Visible heat damage required replacement

of the spring mechanism and if necessary the entire support. We

find that the procedures set forth for the evaluation and restoration

of the safety related HVAC ducts cable trays, cable tray supports,

and the fixed and variable members of pipe supports provide adequate

assurance that these items will be capable of performing as designed

throughout the range of loads and functions specified in the design

bases for Browns Ferry Units 1 and 2.

With respect to the structural steel components and the reinforced

concrete structures which either support the cables or were in the

vicinity of the burning cables TVA conducted a restoration program

which consisted of:

1. Identification of the affected structures;

2. Establishment of the most probable temperature to which each

structure was subjected during the fire;

3. Evaluation of the extent of damage of each structure;

4. Establishment of criteria for the requirement of replacement

or repair of the damaged structure; and

5. Development of procedures for detailed repair.
2-9

The structural steel components affected consisted of miscellaneous

structural steel supports, steel embedment, and portions of the

building superstructure steel. The reinforced concrete structures

affected consisted of walls, columns, and floor slabs, which form

the permanent building superstructure.

The temperature zones were established on the basis of the color of

the structures of different materials of construction, melting of

metals and burning of other materials. The extent of damage of

each structure was evaluated on the basis of the temperature to which

the structure was subjected and the physical condition of the structure.

The criteria for the requirement of replacement or repair of damaged

structures are as follows:

1. Steel Structures

Structural steel components will be replaced or will continue to

be used depending'upon whether their thermal sensitivity is lower

or higher than the temperature to which the structural component

was subjected during the cable fire. In addition any steel structure

found to have been immersed in a temperature zone of 1000OF

or more shall be replaced.

2. Concrete Structures

For reinforced concrete structures, the procedures for the evaluation

of the concrete will consist of detailed visual inspection,

comparison of temperattire zones and testing of samples from

core borings. For reinforcing steel, the procedures of evaluation

will consist of the following:

(a) Analyze the adequacy of the structure by neglecting the steel

bars exposed to high temperatures, and if found adequate

no further evaluation will be done; and,

(b) Conduct sampling and testing of steel bars to determine

their material characteristics. The measured material

characteristics will be evaluated by rechecking the design

of the affected structure. If the reduced characteristics

are still adequate to resist the loads imposed on the

structure, no modifications are necessary.

On the basis of the above criteria, the floor slab at elevation

621 feet in the reactor building which appeared to be most seriously

damaged was evaluated by TVA. Reinforcing steel bar and concrete

core samples were taken and tested. The results of the tests showed
2-10

that the material characteristics of the reinforcing steel and concrete

were not adversely affected by the fire. On the basis of these findings,

TVA concluded that the fire damage to reinforced concrete was

structurally minor. Consequently, the repairs to be undertaken

were considered non-structural.

We have concluded that the criteria, methods and procedures that were

used in the evaluation of damage of structures and in the replacement

of structural steel components assures that all structural components

affected by the fire have been returned to their originally acceptable

condition.

2.4.3 Summary of Inspection Results

The investigations regarding piping, structural and mechanical

components indicated a minimum of damage outside of the cables,

conduit, cable trays and cable tray supports in the immediate

fire area. The piping replaced included only six small nonsafety related lines. Three sections of HVAC duct were replaced.

No structural repairs were necessary. Two spring hangers were

repaired and one pipe support was replaced.
2-11

2.5 Cable Replacement and Repair

The major effort in restoring the facility to a condition in which

it will be able to operate, is the replacement and repair of damaged

wiring and cables. This amounts to some 9500 conductors to be

replaced or spliced. All divisional cabling (reactor protection

system, primary containment isolation system, and engineered safeguards) and 4160 v. power cables will be replaced from terminal to

terminal (without splicing). Only non-divisional control and

instrumentation cables will be considered for splicing. Of these,

the specific repair requirements provide for protection against

high temperature that could result from a poor splice. The procedures

proposed call for splicing at undamaged locations determined by

measurements made on the insulation material; for splices in accordance

with applicable codes; for accessibility to facilitate inspection;

and for measures that assure no mechanical loading exists on the

splices. Appropriate fire stops are provided along the cable trays

away from the splice. In-line splices within the cable trays have

been allowed for those cables that have been demonstrated incapable

of supplying sufficient energy to a poorly constructed splice to cause

cable insulation ignition. All other cables shall be spliced in

metallic enclosures outside the cable trays.

The specific criteria which we and the licensee arrived at for the

replacement and repair of cables are as follows:

1. Reactor protection system, primary containment isolation system,

and engineered safeguards system (divisional) cables shall be

replaced terminal-to-terminal.

2. Damaged nondivisional cables in 4160-volt circuits in cable

trays shall be replaced terminal-to-terminal.

3. Cables in 480-volt circuits shall not be spliced in the cable

spreading room.

4. There shall be no splices within the 15-foot zone containing

trays of both Division I and II identified in Figure 4 of

Appendix 3.3A, Part X, Section A, of the Plan.

5. Cables in medium- and low-level signal cable trays shall be

spliced in-line in the trays without boxes. Medium-level

signal cable trays carry instrument control loop cables,

digital computer cables, shielded annunciator input cables used

with solid-state equipment, and instrument signal cables associated

with transmitters, recorders, and indicators other than thermocouple cables, strain gauge cables, vibration detector cables,
2-12

thermal converter cables, and resistance-type temperature detector

signal cables. Medium- and low-level signal cable trays correspond

to those in items (4) and (5) respectively of paragraph (c) on

page R7.5-11 of the FSAR.

6. Cables in control cable trays shall be spliced in-line in metal

enclosures that meet the requirements of Article 370 of the 1975

National Electric Code. Control rod position cables, which are

medium-level signal cables routed in control cable trays, may be

spliced as permitted by Criterion 5. Control cables correspond

to those of item (3) of paragraph (c) on page R7.5-11 of the

FSAR.

7. Cables in 480-volt trays outside the spreading room shall be

spliced in metal enclosures that meet the requirements of Article

370 of the 1975 National Electric Code.

8. Where cables are terminated on equipment within the reactor

building fire area, and have any portions of the cables routed

in cable trays, the cables shall be replaced from the equipment

such that there is no more than one splice per conductor.

9. Cables routed in conduit within the zone of influence of the

fire area shall be replaced and spliced in the nearest splice

box outside the zone of fire damage containing undamaged cables.

Branch lighting circuits are exempt from this criterion.

10. Spliced cables shall be accessible for inspection.

11. Spliced cables shall be supported so that there is no mechanical

load on the splice.

12. A fire stop, consisting of a Factory Mutual approved fire

retardant cable coating, shall be applied for a distance of

3 feet along the cable trays at both ends of the cable spliced

areas but shall not coat the cable splice.

We concluded that thecriteria used for replacement and repair of

cables were acceptable because they provided protection against the

effects of a high temperature that could be produced in a poor splice

in a circuit carrying a sufficient amount of power and provided protection

against the effects of an interruption to a divisional circuit by

requiring terminal to terminal replacement. We further conclude that

the acceptable restoration criteria in conjunction with the special

training and quality control developed for splicing as well as the

extensive testing of all equipment associated with replaced and

repaired cable provides for an acceptable program for restoration to an

operable condition.
3-1

3.0 Approach for BFNP Fire PrOtection

Units 1 and 2 share a common control room with a shared cable spreading

room located beneath the control room. Cables carrying signals

between the control room and various pieces of equipment in the

plant are routed into the cable trays in the cable spreading room.

The fire caused extensive damage to wiring located immediately

outside of the cable spreading room.

Some sharing of equipment exists between all three units; the

sharing of electrical systems is most extensive in Units 1 and 2.

The electrical design concept employed in the design of Units 1 and

2 is based on using a two-division concept. The purpose of dividing

each Unit's electrical system into two divisions was to assure that

the facility could be maintained in a safe configuration even

with the postulated loss of one entire division.

Consideration of the consequences of fires requires that the

capability be maintained during and following the fire to safely

shutdown the reactor and remove the decay heat. Previous analyses

(PSAR and FSAR) have shown that only one division of electrical

equipment is required to assure that the facility is maintained in

a safe condition. The control rod system in a BWR is designed to

be fail safe. That is, a loss of electrical power results in initiating

a scram of the control rods and shutdown of the reactor. Therefore,

the major objective for protection of the reactor from the effects

of fire is to assure that necessary equipment to remove decay heat

remains operable in the event of a fire.

Thus, it is important to limit fire induced failures to one division

of the safety related equipment for shutdown heat removal from each

reactor which provides for each reactor at least one core spray system

with two associated pumps and valves, and one RHR system with associated

pumps, heat exchangers, and valves, together with enough relief

valves for reactor system blowdown to low pressure. (It should

be noted that non-safety grade equipment is normally used to perform

this function). In order to assure that a fire could not cause a

loss of both divisions, some changes in the facility design and

equipment were necessary.

Although the fire that occurred caused damage to the electrical system

greater than that which had been considered in our original evaluation

of that facility, our current evaluation has shown that even with this

extensive damageconsiderable flexibility remained with respect to

methods available to remove the decay heat from the reactor and

assure the reactor was maintained in a safe condition.
3-2

In spite of the capability already inherent in the facility, the

fire demonstrated that additional emphasis on the protection of the

facility from a fire is needed. In this connection we have considered

those features, actions and design approaches that should be a part

of fire protection capability. The three basic considerations in

minimizing the effects of the fire are: (1) administrative actions

that can prevent a fire from occurring; (2) use of isolation as a

mechanism to prevent a fire from damaging redundant safety

equipment; and (3) incorporation of a means to detect and extinguish

a fire quickly. No one of these three elements can be relied upon

to completely eliminate problems associated with fires and no one

of them can be practically achieved to perfection. However, we

would emphasize that each of these elements should be considered by

itself to the extent practical. The degree of practicality is affected

by the fact that we are assessing the design features, such as physical

separation and installed extinguishing systems, of a plant that

is already designed and built.

In the following sections, the evaluation of these three elements of

the fire protection at Browns Perry are discussed. The significant

modifications made to enhance the fire protection capabilities are

identified.

How these modifications compare with the Nuclear Energy LiabilityProperty Insurance Association (NELPIA) recommendations is presented

in Attachment 3.
4-1

4.0 Plant Fire Protection Reanalysis

The licensee has performed a reanalysis of the total plant fire

protection requirement considering permanently installed

combustible materials, location of these materials in the plant,

identification of the installed fire fighting and monitoring systems,

evaluation of each area to determine if a postulated fire could affect

safe shutdown capability of the reactors, and identified and proposed

modifications to the fire protection and detection systems to ensure

the safe shutdown requirements. The methodology, assumptions and

criteria are set forth in Part X, Sections 5.1.2 and 5.1.2.1 of thb

Plan. The descriptions of the types of fire extinguishing capabilities

used in the evaluation are set forth in Section 5.1.3.

Since the major combustible material involved in the March 22, 1975

fire was cable insulation and jacketing, the licensee made a

concentrated effort on identifying critical areas containing this

material and in addition analyzed all other areas of the total

plant for combustible material and determined their relationship

to the safe plant shutdown requirement.

In performing the analysis the licensee developed a table indicating

the location of combustible materials, the quantity of material

present, the class of fire it presents, the fire detection and

extinguishing methods that are or will be provided for each area

containing combustible material, the resulting fire load that the

material can place on the area, location of the specific areas as

related to plant layout drawings, and a safety classification for

each area, such as safety related, not safety related and safety

related exposure fire potential. This Table is presented in Part

X of the Plan and identified as Table 5.1-1, Sheets I through 9 with

three attachments. The location of each area is shown on drawings

following the table.

The results of the reanalysis are set forth in Part X, Section

5.1.5 and the design modifications are described in Sections 5.2 and

5.3. Our evaluation of these modifications is presented in later

sections of this report.

The basic approach to the fire hazard reanalysis was to survey the

plant and determine the combustible materials present and then

calculate the fire load that they presented to the area. If critical

areas that could affect safe shutdown were involved then the fire

load was used to determine the adequacy of installed monitoring and

fire extinguishment systems. If the installed systems were not

adequate based on the fire load for the area, modifications to the

plant were proposed to protect the critical areas.
4-2

An example of the licensee's approach is for the area where the

March 22, 1975 fire occurred. The cable insulation and jacketing

for this area was determined to be 228 cubic feet of high density

combustibles and 417 cubic feet total. The fire load was calculated

assuming that the fire retardant coating that will be installed on

the cables was not present and was only taken credit for in retarding

the rate of spread of the postulated fire in the area. The calculated

fire load for the high density area was 58,160 BTU per square foot

of floor area and the remainder of the total was calculated to be

15,900 BTU per square foot of floor area. Based on the fire load

the licensee decided to install a fire water deluge system for the

high density area and a preaction sprinkler system for the remainder

of the area. Using the fire load values the proposed systems were

designed to NFPA Standards. The resultant fire water application

was determined to be 0.3 gpm per square foot of projected area of

the cable trays in the high density area and 0.3 gpm per square foot

per 5000 square feet of floor area for the sprinkler system.

The other areas of the plant with installed fire extinguishment systems

were evaluated in the same way. Where fire hoses and portable extinguishers

are the primary fire extinguishing method, total BTU fire load has

been calculated and evaluated against the extinguishing capability.

Our review of these areas indicates that the areas noted are not

related to safe shutdown of the reactors and that provisions such

as diking, drain pits or physical separation would not permit the

spread of these fires to critical areas.

It should be noted that as a result of the fire hazard reanalysis

the licensee expanded the smoke and heat detection system throughout

the plant.

We have reviewed the fire hazard reanalysis presented in Part X

of the Plan and have also made site visits to survey some of the

critical areas described by the licensee. Based on our review, we

conclude that the licensee has defined and located the critical areas

of the plant that could be affected by a postulated fire and potentially

affect safe shutdown of the reactors. The analysis is considered

conservative since no credit was taken for the fire retarding properties

of the coating material that will be applied to all cable trays located

in critical areas. In addition, the design modifications that resulted

from the reanalysis are reasonable and will provide a defense in depth

to the fire detection and extinguishment for the plant when coupled

with the new and revised administrative procedures for fire prevention,

the revised fire brigade training and other design changes and

procedures that are directly related to fire protection.
5-1

5.0 Fire Protection System Impairment Analysis

The licensee has performed an impairment analysis of the high

pressure fire water system and the CO2 system as they will be

installed at the time of restart of Units 1 and 2. Any other

modifications to the systems that will be made at the first refueling

will nave included in their design considerations, potential single

impairments which could render the system inoperative.

The guidelines for the analysis are set forth in Part X, Section

5.4.1 of the Plan and the results are described in Section 5.4.3.

Table 5.4-1 lists the impairments considered, the system impaired,

the specific potential consequences, and the proposed correction.

The proposed modifications to correct the problems identified

are described in Section 5.4.4.

The licensee's analysis has included the potential impairments that

were determined from our review of the fire system process diagrams

and from detailed discussions with the licensee. As a result the

following modifications will be incorporated in the systems, prior

to restart or as noted. Our evaluation of each is included:

1. The isolation valve in the discharge line from the pump header

(valve 0-26-528) will be removed from the system to preclude

a potential impairment of loss of pump flow to the yard loop.

We agree with the licensee that this valve is not required

since other yard loop isolation valves are installed and can

serve the same function. Therefore, we conclude the modification

is acceptable.

2. The electrical power supplies for the three fire pumps are run

in a common cable tunnel and if postulated events occurred, such

as a fire in the tunnel, all the fire pumps could be lost. The

licensee proposes to install a diesel-engine driven fire pump

at another location in the yard loop and take its water from

another source other than the pumping station intake structure.

The addition of this pump will provide an alternate fire water

source and pumping capacity of a minimum of 2600 gallons per

minute to the yard loop. This capacity is adequate to provide all

water needs for fire extinguishment in all areas critical to safe

shutdown of the plant. The licensee plans to add the diesel driven

pump at some time after restart of Units 1 and 2 and, therefore, will

develop, implement and test a casualty procedure for running

temporary power cables from the Unit 3 shutdown boards to the fire

pumps and reestablishing the pump operation within two hours.
5-2

We have reviewed the proposed modification of the addition of

the diesel driven pump and conclude that when installed it will

provide an alternate fire water source and pumping capability

to maintain an operational high pressure fire water system for

the critical areas of the plant required for safe shutdown. The

casualty procedure proposed, to provide temporary power to the

installed fire pumps appears reasonable on the basis that only

certain events can be reasonably postulated that would cause

loss of the normal power supply. The one event that appears the

most probable is a fire in the cable tunnel. With the current

design of the fire water system, the raw service water system

would provide the required fire water to extinguish a fire in

the tunnel. The casualty procedure would then reestablish high

pressure fire water protection for the plant within a two hour

period. The requirement for reactor shutdown has been included

in the proposed technical specifications for each unit of the plant

in the event of loss of the fire pumps. We conclude that, for

the interim period between restart and installation of the pump,

the casualty procedure is an acceptable method to maintain fire

protection in the event of this postulated impairment. Since

the fire that may result in a loss of the fire pump power cables

would be wholly confined to areas not relying for fire protection

solely on high pressure fire water, we find the two hours that

the plant may not have full fire water pressure to be acceptable.

The Technical Specifications for the plant will require that

in the event of this impairment prompt action shall be taken to

restore the fire pump capability or to place the reactor in a

standby condition.

3. The fire hose racks for the Diesel Generator Buildings were

determined by the licensee to have a potential impairment which

would prevent fire water being available to the areas they serve

by a sectionalizing valve being closed. The licensee proposes

to modify the hose rack water supply system by the addition of a

redundant fire water line from the reactor building fire water

loop. These hose racks will be relied upon in the interim

period to provide fire fighting capability in the pipe and

electrical tunnel of the buildings.

The primary purpose of these hose racks is to extinguish small

postulated fires in this area, and if required, to cool the fire

door of a diesel generator compartment in the event the automatic

CO2 system in the diesel compartment should be impaired at the

time of a fire in the compartment.

We conclude the modification to add a redundant fire water line

to the hose racks will eliminate the original impairment and is,

therefore, acceptable.
5-3

4. The impairment of the manual, electrically actuated CO2 system

valves for the cable spreading rooms is covered by the fire brigade

using the manual valve provided to discharge the CO2 system

into the cable spreading rooms. As a back-up fire hoses are

installed in the control building, located such.that they can

be brought into any of the spreading room doors.

Considering the redesign of the spreading rooms' fire stops, the

application of Flamemastic to all cable trays in the rooms, the

addition of smoke detectors in the rooms and the administrative

procedures to maintain the rooms free of combustible materials,

we believe that manual fire fighting means are adequate for this

impairment. Test data included in Part X of the Plan indicates

that fires originating from cable trays coated with Flamemastic

will not spread and will be contained to a local area.

5. The impairment of automatic operation of the CO2 system serving

the Diesel Generator Building electrical board rooms for the

two Diesel Generator Buildings could result in the loss of one

of the electrical board rooms. The licensee has proposed to

install an automatic fire door between each board room that will

close on actuation of either of two smoke detectors that are

placed on each side of the door at ceiling level. The CO2

system can be operated by manual actuation of the valve at

the local control station. If this cannot be accomplished

the fire door will protect the redundant electrical board

room from the postulated fire. The fire door is considered

adequate to contain the fire resulting from the fire load in

each electrical board room.

Based on our review of the Diesel Building area, the fire

extinguishing systems available and the addition of the fire

door; we conclude that the postulated impairment would not

jeopardize the safe shutdown of the plant.

6. The impairment of the valve in the discharge line of the head

tank, connected to the fire water system, (Valve 25-32) by

failing open would result in a drop of pressure and flow rate

in the fire water system when the high pressure fire pumps

were turned on. Therefore, the performance of the high pressure

fire water system could be degraded to a point where required
5-4

flows and pressures would not be available to the plant. The

licensee proposes to modify an existing series valve (Valve

25-720) to be a remotely operated isolation valve. This modification would then result in two valves in series to isolate

the head tank from the fire water system when required.

Our review of the modification indicates that the addition of

the second remotely operated valve will eliminate the original

postulated impairment and, therefore, we conclude it is acceptable.
6-1

6.0 Administrative Changes

6.1 Training (Fire Related)

Since the March 22, 1975 fire, the licensee has re-evaluated the

training provided to the operations personnel to determine what

changes, additions or new requirements should be incorporated

in the training programs. Based on this re-evaluation the

training program has been modified. These include

revision of the Fire Brigade Leader Training Course, the

Fire Brigade Member Refresher Training Course, the Fire Brigade

Member Training Course and the Student Generating Plant Operator

Training Program to include added training on the use of water

to extinguish electrical fires and/or fires in cables, cable

tunnels and cable trays through the use of water fog; the proper

use of self-contained breathing apparatus; the proper procedure

to recharge the breathing apparatus air cylinders; the use of

ventilation systems for smoke control; and the relationship between

the Fire Brigade Leader, the Shift Engineer and any off-site fire

fighting personnel that may be called to assist the Fire Brigade.

Since the fire a number of the Shift Engineers and Assistant Shift

Engineers have completed the revised Fire Brigade Leader Training

Course. In addition, a Fire Brigade Member Refresher Training

Course has been given to all operating plant personnel that could

be assigned to the Fire Brigade.

Periodic drills will be conducted at the plant and evaluated by

the plant supervision and members of the Safety Engineering Services

Office of the Division of Power Production. These drills will

include use of the pre-fire plans that will be developed for the

critical areas of the plant required for safe plant shutdown.

In addition, the Fire Brigade will be familiar with and practice

the casualty procedures.

Implementation of a formal indoctrination program has been

instituted to familiarize construction workers and contractors that

may perform work in operational areas at the plant. This will

include personnel that have received the plant and security orientation.

This program will emphasize fire reporting and emergency procedures.

Our review of the deficiencies that were reported as a result of

the March 22, 1975 fire indicates that the licensee has made adequate

modifications-or revisions to the formal training programs that

should minimize future difficulties. In addition, training and drills
6-2

that will be accomplished for the unique features of the Browns

Ferry Nuclear Plant will assist in providing a well organized "

Fire Brigade function. This, in conjunction with the procedural

changes being implemented for fire prevention, fire watches, control

of temporary combustible material, work plan control and the pre-fire

plans, provides the basis for our conclusion that the training of

personnel in the operating plant is acceptable for prompt and effective

Fire Brigade and reactor operator actions to extinguish fires and place

the reactors in a safe shutdown condition.
6-3

6.2 Procedures CFire Related)

As a result of the March 22, 1975 fire, the licensee has revised or

developed improved procedures'aimed at correcting or preventing

conditions from developing that could become 8 fire potential. In

addition, as a result of the design modifications and additions,

new procedures have been developed to implement actions that are

required to provide appropriate fire prevention, fire fighting,

operational requirements, maintenance and testing and other fire

protection related measures.

A number of these procedures are classified as fire prevention

procedures. They include the following:

1. Cutting, welding and open flame work permit control, which

requires that potential fire considerations must be made by

the authorizing supervisor and a determination made of the

fire fighting equipment available in the area of work and

whether a fire watch is required;

2. Procedures for control of all fire related testing, maintenance,

record keeping and assistance to the various operations groups

by an individual assigned to these specific duties;

3. Procedures for performing the required maintenance and testing

of the fire protection system and equipment. This procedure

designates the responsible plant staff individual for this

procedure;

4. Procedure for valve supervision program and record as related

to the fire water and CO2 systems, including the yard loop and

pumping station;

5. Procedures for the requirements when a fire stop is breached

for any purpose. These include the requirement for a fire

watch at the location until the fire stop has been restored

to its original design condition;

6. Procedures for housekeeping to reduce or eliminate combustible

materials from areas required for safe plant shutdown;

7. Procedures requiring a roving fire watch during the interim

between restart and the completion of all fire protection

modifications. This procedure will require clock registers

be used by the fire watch at points in the plant that are critical

to safe plant shutdown. The inspection interval for the clock

register stations will be every two hours;
6-4

8. Procedure for controlling flammable liquid storage in the

plant and requirements for the type of containers that can

be used when such liquids are required for plant operation.

These containers are small and are Underwriters Laboratory

approved;

9. Procedures for a formal plant self-inspection program for fire

safety. This Procedure Will require plant supervisory inspections

using prepared fire safety checklists on a periodic interval;

10. Procedure for assessing transient fire loads that are required

for plant operation and maintenance. This procedure is being

developed to reduce and control temporary fire loads in the plant

that could present a potential fire hazard to critical areas of

the plant required for safe plant shutdown. The development of

this procedure is in progress and the plant supervision is being

assisted by TVA's Safety Engineering Services to establish

guidelines for use of the plant operating staff. These guidelines are to be based on the maximum temporary fire load that

can be located in a critical area and be protected by the existing

or future fire extinguishing systems. If the temporary fire

load exceeds this capability then the guidelines will state

the additional fire protection required while the temporary

loading is in the plant. We have concerns in this Procedural

area that we believe must be considered in developing the

guidelines. The major concern is that these temporary materials

may be located in a critical area and credit taken for a fire

extinguishing system that may not be effective where the materials

are located. Specifically, areas that contain critical systems

or cabling that are protected by a deluge or sprinkler system,

but because of cable tray, ventilation ducts or piping obstructions

located between the fire extinguishing system and the location

of the temporary material they may not be covered by the systems

and therefore present an exposure fire potential to the critical

area. We will require a continuous fire watch with charged

hoses available until the temporary materials are removed from

such areas;

11. Procedures for outside or independent fire protection/loss

prevention inspection program to be performed on a periodic

basis. This is a revision to a current procedure and will

require an independent inspection by utilizing either proper

TVA personnel or an outside fire protection firm. This inspection

program is to be instituted and the first inspection performed

prior to restart of the reactors and annually thereafter.

This inspection is in addition to the plant self-inspection

procedure.
6-5

Other procedures are being revised or will be developed to provide

for written and approved controls for fire protection requirements.

These include the following:

1. Procedures for assuring maintenance of the pressure-seal fire

stops by periodic leak detection tests at the penetrations.

This procedure will provide a record of the potential deterioration

or leakage of the fire stops and provide a basis for corrective

action;

2. Procedures for reestablishing high pressure fire water to the

yard loop if power is lost to the fire pumps. This procedure

will include the requirement for temporary connections from

the Unit 3 shutdown boards and the connections will be fabricated

and tested prior to plant startup. In addition, the procedure

will include measures to be taken, if needed, to bring in

backup fire truck pumpers to establish fire water to the yard

loop;

3. Procedures for the use of existing ventilation systems for

smoke control. This procedure will include instructions on

the remote manual or manual use of ventilation systems that

may be used for smoke control. The automatic capabilities

for isolation for all isolation signals will be maintained.

4. Procedures that will include pre-fire plans for the critical

areas of the plant required for safe reactor shutdown. These

pre-fire plans will be incorporated into the fire brigade in-house

drills to provide a direct approach to a potential fire area.

They will also include plans for providing recharged breathing

air cylinders to the fire brigade;

5. Procedures for semi-annual chemical treatment of the fire

water system, periodic flushing to assure clear piping and

annual inspections for removal of crustacean accumulation

where needed. This procedure is being added to existing

procedures for crustacean control of all water systems;

6. Update procedure for providing supplemental fire protection

measures when an installed system, including monitoring, is

taken out of service for maintenance, repair or testing. This

procedure requires a sign-off method for getting approval to

cause the impairment and to inform the designated responsible

individual that the impairment has been corrected and the

system has been placed in service.
6-6

We have made an overall review of the procedure outlines and

commitments made by the licensee to update, revise or develop

procedures for fire prevention and for fire protection requirements.

These procedures are being coordinated by the responsible Divisions

of the TVA organization. In addition, our review considered the

organizational changes that will result in a plant-wide review

of the fire related procedures being developed and implemented.

Based on this review and our requirement as stated in Item 10, we

conclude that the licensee is taking the required actions to provide

assurance that adequate administrative controls are being implemented.
6-7

6.3 Organization (Fire Related)

TVA has made some organizational changes as a result of the fire that

increase significantly the attention that will be given to fire

protection in the future. A Fire Protection and Prevention Board

has been formed with representatives from the Division of Engineering

Design, Division of Power Production and the Division of Construction.

This Panel will provide coordination between these divisions for the

plans and policies concerning fire protection at all TVA power plants.

TVA has hired two fire protection engineers to work under the

Supervisor Safety Engineering Services in the Division of Power

Production. They are temporarily assigned to Browns Ferry until

the Engineering Aid (Safety) position has been filled at the plant

Their permanent duties in the Safety Engineering Services group

include conducting fire training and drills for plant personnel,

reviewing and developing fire preplans for emergency situations,

reviewing and evaluating work practices at the plants, and conducting

periodic inspections and audits at TVA plants.

At Browns Ferry, the Plant Superintendent is responsible for implementing the Division of Power Production procedures for fire

protection and prevention, and for developing operator procedures

for fire emergencies and fire protection system operations.

The BFNP Operations Supervisor has been given the responsibility

to review all work packages for fire hazards and to identify controls

needed before approval of the work. An Engineering Aid (Safety)

position has been established and is temporarily being filled by

the Safety Engineering Services engineers as mentioned above.

The duties of this position include routine in-plant fire inspections

and maintenance of all fire equipment in an operable condition.

The Fire Brigade at the plant has been reorganized. The Brigade

Leader is now the Assistant Shift Engineer for Unit 2 instead of the

Shift Engineer. Two Assistant Unit Operators from each of Units

1 and 2 are assigned as Brigade members. The Shift Engineer no

longer has the dual responsibility of safe operation or shutdown

of the plant and directly leading the Fire Brigade. Specific

assistant unit operators are now assigned to the Brigade so that other

operators may concentrate on plant operation or shutdown.

Finally, a position of Restoration Coordinator has been created

that will remain in effect until all the restoration work, including

the modifications to be completed in the future, are completed.

He will have available to him expertise in fire protection methods

and system design. Responsibilities of the Restoration Coordinator

include the following:
6-8

1. Principal coordinator for all design, construction, and operational

activities relative to restoration of Units 1 and 2,

2. Review and approval of all documents submitted by TVA to NRC in

connection with restoration, and fire protection and prevention

improvements,

3. Overall planning, establishment, and maintenance of a critical

path schedule for restoration and completion of modifications,

and

4. Coordination and approval of TVA's overall efforts in fire

protection and prevention improvements, including design and

installation of new systems and changes necessary in fire fighting

methods and techniques.

We have concluded that the above organizational changes and additions

provide an improved degree of management in the fire protection aspects

of plant safety from the highest level of management down to the

working level for fire prevention, fire fighting, fire protection

system design, training, testing, inspections and audits. We have

concluded that these organizational changes are acceptable.
6-9

6.4 Quality Assurance

TVA's Quality Assurance Program for the restoration activities

at Browns Ferry Nuclear Plant Units 1 and 2 has been previously

reviewed and found acceptable prior to the restoration work as

described in the Safety Evaluation Report issued with Amendments

10 to DPR-33 and 7 to DPR-52 on June 13, 1975 authorizing the

commencement of clean-up and restoration following unloading of

fuel.

The licensee has described the Quality Assurance Program For

Station Operation in Section D.4 of Appendix D to the FSAR in

Amendment 62. In Appendix D.4A, the licensee commits to follow

the guidance contained in the WASH documents 1283 (5/24/74) "Guidance

on Quality Assurance Requirements During Design and Procurement

Phase of Nuclear Power Plants - Revision 1", 1284 (10/26/73)

"Guidance on Quality Assurance Requirements During the Operations

Phase of Nuclear Power Plants", and 1309 (5/10/74) "Guidance on

Quality Assurance Requirements During the Construction Phase of

Nuclear Power Plants" based on his understanding that the term

"shall" denotes a requirement and the term "should" denotes a

recommendation.

We find this general exclusion unacceptable as stated and will

require that each recommendation, "should", that is not being

followed be specifically identified by TVA in the QA Program and

justification provided for our evaluation. We will resolve this

open item prior to authorizing return to operation and will include

our evaluation of the QA-Program in a Supplement to this report.

7-1

7.0 Design Modifications

7.1 Electrical Changes

In addition to replacement and repair of damaged wiring and cables,

TVA proposed design changes to enhance fire protection and eliminate

the source of the loss of redundant safety equipment which occurred

as a result of the March 22, 1975 fire.

The most significant loss of redundant equipment was associated

with failures of their power sources which were caused by short

circuits to lamp circuits leading from the control circuits of

Reactor Motor Operated Valve (MOV) boards. These cables were

originally considered to be non-divisional because a dropping

resistor was provided in the lamp circuit for the purpose of isolating

this circuit from the control circuit. Several of these cables

from both divisions were therefore allowed to be run together in

common trays and were, subsequently, damaged by the fire. TVA has

proposed to remove the cables leading from Reactor MOV board control

circuits to breaker indication lamps to eliminate the major problem

which was the loss of ability to operate the boards in both divisions.

We agree with this approach and conclude that these circuit changes

will eliminate the loss of redundant division MOV board controls.

The second most significant cause for loss of redundant equipment

was due to the proximity of conduits containing division I and

division II cables to cable trays which were the primary source

of combustible material. Although the use of conduit may provide

adequate protection against some effects of fire and other hazards

for a short time, the fire which occurred proved that conduit

protection alone was insufficient for a fire of that magnitude without

additional protection in the form of additional separation or fire

barriers. To remedy this TVA is modifying the cabling and wiring

systems. The modifications include physical moving of certain

conduits, use of Flamemastic to provide an effective barrier,

and the addition of fixed fire water systems.

TVA's analysis of the plant, as described in Part X of the

Plan, indicated 90 conduits associated with the three units which

ran parallel to cable trays within their established "move zone".

Conduits crossing cable trays were not considered in the analysis

on the basis of it being impractical to move those conduits. Of

the 90 conduits identified, 44 are being relocated. This will be

accomplished prior to restart. Considerations leading to the
7-2

judgement that conduit would not be relocated included physical

limitations and the potential for creation of new divisional interactions. The use of Flamemastic as a barrier is discussed in the

following section. Although there are locations in which conduit

crosses or runs parallel to cabling of another division, the limited

area involved where conduit crosses cables, the thorough protection

provided by the Flamemastic coating, and the enhanced fire detection

and fighting capability of the facility lead to our conclusion that

all identified potential cross divisional fire effects due to

proximity of conduits have been eliminated. The additional fixed

fire water system which will be installed at the first refueling

outage, will be specifically designed to ensure that divisional

conduit will be thoroughly wetted where both divisons could be

affected. This automatic system when installed will enhance the

fire extinguishing capability provided by the fire brigade.

In evaluating the TVA modifications concerning divisional isolation

we recognize that complete physical independence cannot be achieved.

We conclude that the TVA approach which includes the complete coating

of all exposed cables with Flamemastic and the addition of the fixed,

automatic fire water system to cover all areas of possible cross

divisional fire effects provides sufficient protection against loss

of more than one division of safety equipment in the event of a postulated

fire.

A third cause for loss of redundant equipment was damage to their

cables sharing the same cable tray contrary to the stated criteria.

This third significant cause for loss of equipment will be remedied

by separating the involved cables. Selective changes are also being

made in the electric power system to improve isolation. These

changes will provide individual normal feeders to the 4KV/480-V

transformers. One change will eliminate sharing between units of

the 4KV feeder that was the normal supply to 480-V Shutdown Board

2A and 2B through TS3E. Other changes will provide individual

4KV feeds to transformers TDA and TDB which supply 480-V Diesel

Auxiliary Boards A and B, respectively. The only shared 4KV feeder

will be the alternate supply to transformers TSIE and TDE. These

changes will be accomplished prior to restart. We have reviewed

these changes and find that they are acceptable.
7-3

7.2 Use of Flamemastic 7lA (Fire retardant coating)

The licensee proposes to use Flamemastic 7lA, (described to be a

non-flammable, non-toxic, odorless, waterbase compound consisting

of thermo-plastic resinous binders, flame retardant chemicals,

and inorganic non-combustible reinforcing fibers), to provide a

fire retardant barrier on all exposed cable in the secondary

containment area of the reactor building, cable spreading room, diesel

generator building, the water pumping station, and the pumping

station cable tunnel. These are the only areas containing safety

related equipment necessary for safe shutdown. The coating is to

be applied to all cable types, including non-divisional cabling, to

a wet thickness of 1/4 inch to effect at least a 1/8 inch dry

cured thickness.. This thickness of coating has been demonstrated by

tests to be more than adequate for the application. The manufacturer

recommends a 1/8 inch wet thickness for this type of application.

This coating will greatly reduce the potential for a fire and will

also limit the propagation of a postulated fire.

A number of tests that have been performed to determine the effectiveness

of Flamemastic 71A as a fire retarding barrier for electrical cabling,

similar to or the same as utilized in the Browns Ferry Nuclear Plant,

and to determine its capability to retard propagation of a fire.

Some of the tests were performed using short time period (up to

15 minutes) fire initiators, such as, gasoline, naphtha, gas, oil

and transformer oil. The results from these tests demonstrated

that, for the test fires, Flamemastic 71A would protect the cabling

and effectively prevented propagation of the fire to the coated

cables. Longer time fire tests (up to 2 hours) were performed by

TVA where the fire initiator was a gas burner with a flame temperature

of 1500OF and exposed cable insulation was available as a fuel

contributor. These tests demonstrated the effectiveness of Flamemastic

71A to stop propagation of a fire from burning cables and also

demonstrated that for temperatures of 1SO0OF the Flamemastic 71A

protected the cable and a fire could not be initiated. In the degraded

test performed by TVA, where the fire stop had been dismantled up

to the penetration sleeve, it required the Flamemastic 71A to be

broken up and additional combustible materials added, in addition

to the gas burner, to develop a fire for testing the penetration

seal. This provides further evidence that the coating is fire

retardent and will provide an effective barrier to fire for an extended

period of time. For the types of fires that could be postulated to

occur in critical areas in the Browns Ferry Nuclear Plant, temperatures

in the same range as the test temperatures are expected.

Based on the test results from the Watts Bar cable tray fire tests,

and from the other tests performed and included in Part X of the

Plan, we find that Flamemastic 71A is an effective barrier against

the initiation and propagation of fires and is a significant
7-4

improvement over the one hour thermal barrier originally proposed

by TVA. The use of the coating in conjunction with other fire

extinguishing systems, discussed in later sections of this report,

will provide the defense in depth protection for the essential

divisional cabling that controls or powers the equipment required for

safe shutdown of the reactors.

Part X of the Plan contains the results of ampacity tests utilizing

a Flamemastic 71A coating on the cables to determine appropriate

derating factors. Ampacity of the installed cables has been reviewed

and based on the test results, a reduction of 1 to 10 percent in

ampacity of the coated cables is indicated. TVA has utilized the

maximum reduction of 10 percent as a criterion for all coated cables.

A study of power cables in trays shows that a 10 percent reduction

in the ampacity of the cables would still allow maximum design

current to be carried with an adequate margin.
7-5

7.3 Cable Spreading Room Ventilation

The licensee proposes to install two dampers in the exhaust ducts

from Unit 1 and 2 cable spreading room and Unit 3 cable spreading

room. This modification was required to provide separation between

the cable spreading rooms where a common exhaust system is installed.

The new dampers are installed in each exhaust duct from each spreading

room upstream of the common exhaust duct. Additional dampers are

installed in the discharge ducts of the two exhaust fans serving the

spreading rooms. Therefore, with actuation of the CO2 system in

either spreading room the room will be isolated automatically.

The control and power circuits for the ventilation system are being

rerouted through areas outside their respective spreading rooms and

therefore provide increased assurance that the ventilation system will

be automatically isolated upon manual actuation of the CO2 system.

In view of the additional dampers and the new penetration-seal fire

stop design, TVA re-evaluated the cable spreading rooms to assure

that when the CO2 system is discharged, the rooms would not be

pressurized above the tested pressure of the penetration-seal fire

stops. As a result of the re-evaluation, TVA proposed to install

a counter weighted back draft damper in the spreading room wall

common with the turbine building. This damper would release a

pressure buildup above five inches of water, which is the pressure

to which the fire stops have been tested, and vent the room until

the CO2 system has discharged the amount of CO2 to reach the design

concentration in the room. The back draft damper would then close

to maintain the CO2 concentration.

Based on our review of the proposed changes to the cable spreading

room ventilation system, we conclude that the licensee has now

provided for better isolation of the room to enable proper

operation of the C02 system.
7-6

7.4 Fire Stop Designs

The licensee has proposed a new, improved penetration seal - fire

stop design for use wherever the secondary containment function

for the reactor must be maintained. The penetration seal - fire

stop design function is to provide for the maintenance of at least

a 1/2 inch of water negative pressure differential between the

secondary containment boundary and the environment and to provide

a fire stop.

The proposed design uses a silicone room-temperature-vulcanizing

(RTV) foam material to make the seal around the electrical cables

and conduit that pass through the secondary containment boundary.

Based on tests performed by the licensee, the seal material cannot

be classified as non-combustible and therefore must be protected

by other materials to retard the spread of a postulated fire that

could contact the seal material. The licensee's Watts Bar Tests

showed that fires completely consuming the cabling materials on

one side of the fire stop did not affect the seal. Burning periods

in excess of one hour at the face of the penetration seal-fire

stop were involved.

No direct comparison can be made between the flammability and/or

fire resistance as related to full scale fire conditions for the

original polyurethane seal material and the silicone RTV compound

based on standard testing methods. These testing methods are limited

to laboratory type tests and as such are extremely limited in their

ability to confirm the uses of these materials in their installed

configuration. However, general experience in the use of polyurethane

and silicone RTV compounds in industrial applications has indicated

that while polyurethane has shown itself to present an unacceptable fire

potential in many cases, the widespread industrial use of silicone RTV has

not been identified as presenting any particular fire problems. Thus

general experience coupled with the licensee's Watts Bar full-scale tests,

discussed below, indicates the acceptability of the silicone RTV

compound as an acceptable material when utilized in accordance with

the proposed design.

The proposed penetration seal-fire stop design provides the fire

retarding and fire resistant materials that protect the seal material.

The proposed design is described in Part X, Section 4.1.1 of the Plan

and shown in Figures 4.1-1 through 4.1-4. Through the use of the

combined fire resistant materials (Cera-Form, Cera-Felt, Cera-Fiber

and Cera-Kote) and Flamemastic 71A (the fire retardant coating material),

the silicone RTV seal material is protected. The testing performed

at the licensee's test facility at Watts Bar confirms that for the

types of cable fires used in the tests, the penetration seal - fire

stop is acceptable for its intended function when used in the

configuration specified by the proposed design.
7-7

Other fire stop designs are utilized throughout the plant that

are not required to perform a pressure-seal function. The licensee

has proposed a design for this application and it is described in

Part X, Section 4.1.3 of the Plan. It consists of a fire resistant

board installed on each side of the penetration with the void between

the boards being filled with an inorganic fiber for a maximum depth

of up to one foot. The entire exterior surface of the assembly

will be coated with Flamemastic 71A. Based on test data provided

by TVA, tests performed by others that used the silicone RTV compound

instead of the inorganic fiber and no Flamemastic coating, and

preliminary results from a test performed on the type of fire stop

proposed by TVA, we find that the proposed design is acceptable

and will delay the spread of a postulated fire through the penetration for a sufficient period of time to permit extinguishment of the fire

in the areas utilizing this design.

We therefore conclude that, with a sound quality assurance program

to assure that the designs are installed in the plant as described,

the penetration seal - fire stop and the non-pressure seal - fire stop

designs are acceptable for use in the Browns Ferry Nuclear Plant.
7-8

7.5 Design Changes Related to Fire Detection and Extinguishment that are

to be Completed Prior to Startup

7.5.1 Fixed Water Systems

As a result of the reanalysis of the fire hazards that could result

from combustible materials that are a permanent part of the plant,

the licensee has proposed several changes or modifications to the

fixed high pressure fire water system for each unit. We do require

additional information in regard to TVA's consultant's disagreement

with the hydraulic calculations for the fire water system. (See

Attachment 2) These changes and modifications are set forth below

and include our evaluation of each.

1. Provisions will have been made to permit fire pump performance

tests at regular intervals, the pressure which the high pressure

fire pumps will maintain on the water system will have been

increased from 130 psi to 175 psi and the high pressure piping

that has not been hydro-tested to 225 psi will have been so tested.

These three items represent changes to the existing system

that resulted from the reanalysis. We agree with the licensee

that these changes are necessary and acceptable to assure that

the high pressure fire water system can perform its design

function.

2. Additional zoned fixed deluge systems will be installed in

areas where a high concentration of cable trays exist and where

access to these areas is restricted due to congestion of trays.

These areas are the North Wall of the Reactor Buildings on

Elevation 593.0 ft.,the area at column line R2 and P on

Elevation 565.0 ft.,and all penetrations from the cable

spreading rooms into the Turbine Building. The design basis

spray rate for these systems is 0.3 gpm/ft 2 of projected area

of the cable tray stack. This rate was based on the fire

load discussed in the fire hazard reanalysis of this report.

Consideration was given to the existing building drainage

capability in establishing the zone areas. Consideration

was also given to location of safety related components in

the areas covered by the deluge system. We conclude that

no safety related equipment will be affected by operation

of the deluge system.

For the restart of Units 1 and 2 these systems will be manually

actuated at locations in the area of the deluge systems. Smoke

and heat detectors will alarm in the control room and locally

for each zone to permit the prompt dispatch of the fire brigade

to the area. Other areas of the plant will be protected by the

Flamemastic coating, expanded smoke detection, and manual fire

fighting during the interim between restart and the first

refueling of the Unit.
7-9

Based on our review of the deluge systems and considering the

addition of smoke detection systems (Sec. 7.5.3), augmented

procedures (including pre-fire plans), testing, fire brigade

training, and the requirement for a roving fire watch who will

monitor all critical areas at least every two hours until these

systems are automated, we conclude that the manually operated

zoned deluge systems provide an acceptable fire protection

capability for these areas.

3. The reanalysis indicated that certain areas of the plant

required additional fire hose racks to provide backup to

existing fire suppression systems and to cover others not

heretofore covered. Additional 1-1/2 inch hose racks will have

been installed at each end of the diesel generator buildings

pipe and electrical tunnels. These racks will have been installed

to cover two areas of the plant not previously provided with high

pressure water hose protection. Three additional 1-1/2 inch hose

racks will have been installed at Elevation 550.0 ft. of the

pumping station. These hoses will provide backup coverage to

the cable tray installation in the pumping station and provide

general coverage for the entire floor of the pumping station.

Three additional 1-1/2 inch hose racks will have been installed

in the control building at Elevation 593.0 ft. and are located

on the turbine building wall side of the building. These hoses

will provide high pressure water for the rooms located at this

elevation and in addition can provide hose coverage through

each door of the cable spreading rooms.

All hoses throughout the plant are neoprene lined hose, with some

of the existing linen hoses being replaced. Hose and nozzle

connections will have been made compatible with equipment used

by the local fire department which in some cases of existing

hose stations will require adapters at the stations to convert

from the Standard Iron Pipe Threads to American National Standard

Fire Hose Threads. In addition, all nozzles are suitable for

use in fighting electrical fires.

Based on the licensee's consultant's recommendation,the hose reels

have been rewound using the double relap methods and preconnected

to the high pressure fire water system. The licensee will also

make a survey of the plant and determine that at least two hose

racks, with appropriate hose lengths, will be capable of providing

water for fire fighting especially for the cable trays and other

areas covered in this section.

Based on our review of the design requirements and the plant

drawings in Part X of the Plan, we conclude that the licensee

has provided the necessary additional hose racks and has

demonstrated that the hose coverage is or will be adequate

for the areas of the plant that are required for safe shutdown

of the reactors in the event of a fire requiring hose protection.

It is recognized that this hose protection plays an important
7-10

role in the fire fighting capabilities during the interim between

restart and the first refueling. During this interim period the

Flamemastic coating will retard cable fires and the smoke detection

system will provide early warning for prompt fire brigade response.

Pre-fire plans for critical areas will also be instituted prior to

return to operation and a roving fire watch is required to monitor

these areas of the plant at least every two hours to ensure prompt

response to any fire. This, combined with the accessibility of

these areas for manual fire-fighting, leads to our conclusion that

these measures are acceptable for the interim period until the

first refueling.

7.5.2 Heat Detection System

Heat detection cables are to be installed on cable trays that will

have the manual zoned fixed deluge system installed and at other

congested cable tray areas as described in Part X, Section 5.3.1.1

and shown on Figures 5.3-1 and 5.3-2 of the Plan. The basic principal

of the heat cable operation is two conducting paths normally separated

by a low melting point salt compound, which when heated to 2500 F,

will melt and form a conductive link and set off the alarm system

connected to it. The proposed system will be installed on a zoned

basis and will cause alarms in the control room and at the local deluge

system valve stations. The indicating system in the control room

will give the local panel location by elevation and coordinates.

The local panel will identify the subzone that is in the alarmed

condition, thus giving the fire brigade detailed location of the

alarm condition.

The system design is such that a break in a heat cable will not

preclude the detector from functioning. In addition, the system

will be monitored and will indicate a power failure in the control room.

The power supply for the system will be provided by the Instrumentation

and Control system bus which has multiple power supplies connected

to it to assure the monitoring systems function even with loss of

off-site power.

We have reviewed the proposed heat detection system for its current

application and conclude that for the congested areas it serves is

acceptable for restart, pending resolution of the concern raised by

TVA's consultant regarding the design and installation requirements.

(See Attachment 2)

7.5.3 Smoke Detection System

The licensee proposes to install an expanded smoke detection system

throughout a number of areas of the plant. The system is described

in Part X, Section 5.3.1.2 and shown on Figures 5.3-3 through 5.3-6.

The additional detectors are the ionization type, that are preset

at the manufacturer's plant to the Underwriters Laboratory

Standard UL167 specified sensitivity of 6 milligrams of

combustion products per cubic foot of air. This sensitivity

is sufficient to detect the smoke from the combustible

materials located within the plant and provide an alarm.

The detectors will be located on a 30 foot grid with some

changes to provide adequate coverage where there are obstructions
7-11

or to provide a more concentrated coverage for certain areas. The

areas are zoned with no less than two detectors per zone. These

zoned areas will alarm in the control room for either smoke or

trouble. Local panels are provided which will give indication of

up to four zones that will indicate the area where the smoke detectors

actuated. The trouble indicators for both the control room and the

local panel will indicate a line break or loss of power to the detector

system in that zone.

Additional smoke detectors are being installed in the cable spreading

rooms to provide adequate detection in air pockets formed by overhead

I-beams and will take into consideration air flow within the rooms.

These detectors plus the originally installed detectors will detect

smoke in the rooms and alarm in the control room. The entire fire

detection system will be maintained and tested in accordance with

National Fire Protection Association (NFPA) Standard No. 72E.

Our review of the smoke detection system that is being added to

the existing system has indicated that all the areas that have been

identified as required for safe plant shutdown will be monitored

by smoke detectors. Detectors for Elevation 621 ft. will be

incorporated into the system prior to exceeding one percent power

level. With this coverage and the annunciation system being provided

in the control room and at local panels, the system will provide

prompt early detection of a fire and enable the fire brigade to take

immediate action to extinguish the fire and is therefore acceptable,

pending resolution of the concern raised by TVA's consultant regarding

the design and installation requirements. (See Attachment 2)

7.5.4 Self-Contained Breathing Equipment

The licensee has reviewed the use of the self-contained breathing

apparatus during the March 22, 1975 fire and based on this review

has made several changes to the requirements for the apparatus.

These changes include increasing the total number of useable apparatus

from 24 to 39 and the number of charged extra air cylinders from

27 to 57. In addition, 6 Chemox Oxygen Breathing Apparatus will be

maintained available for emergency use only.
7-12

Emergency equipment cabinets are located in the control room and

provide for storage of 10 air mask units and 15 extra air cylinders.

The remaining units are stored in the shop area and maintained by

the Health Physics Section.

Due to the improper use of the air mask units during the fire, a

two-hour training course has been conducted for operations employees

to assure thorough familiarity with the proper operation of the

units. In addition, training with air masks is now an integral

part of the-Fire Brigade Training Courses.

The recharging system for the air cylinders has been increased

in volume from eight 800 cubic foot supply cylinders to 16.

These cylinders are used in a cascade charging system. Based on

the review performed by the licensee of the deficiencies in the

use of the charging system a procedure for recharging the air mask

cylinders has been issued and the operations personnel trained

accordingly.

In order to provide essentially an unlimited air recharging capability,

the licensee has installed a high pressure air compressor which has

the capacity of providing breathing air at 20 cubic feet per minute

at 2500-3000 psi pressure. An air mask air cyclinder requires 45

cubic feet of air at 2200 psi pressure to be considered full.

We have reviewed the events that were reported for the deficiencies

of the self-contained breathing apparatus and the modifications and

additions to the breathing air overall storage and supply that now

exists at the plant. In addition, we have reviewed the procedure

outline that will be incorporated into the plant operation, which

includes the charging system, pre-fire planning that will incorporate

a system for providing charged cylinders to the fire brigade, and the

number of units available for immediate use with respect to the number

of personnel required for the fire brigade. Based on our review, we

conclude that the self-contained breathing equipment as currently

installed in the plant is adequate for use of the fire brigade and

control room personnel for a postulated fire.

7.5.5 Communication System

The plant communications systems that were available at the time

of the March 22, 1975 fire consisted of (1) the main Bell Telephone

system, (2) the Private Auxiliary Exchange (PAX) System that provides

for an internal plant-wide telephone system and also connects to

other TVA communications systems, and (3) Plant Wide Public Address

System which provides for immediate dissemination of information and

instructions into all areas of the plant and is used for directing
7-13

fire brigade and support personnel activities.

A portable Radio Emergency Command Net has been added to the plant

communications systems and tested to verify satisfactory plant

coverage. The system provides for three 5-watt portable radios that

are assigned to the plant emergency director, the fire brigade

leader and the fire department chief (if required). Using this

system in conjunction with the other communication systems described

above, the plant has an assured means of communication during

emergencies, including fires.

We conclude that with the addition of the radio system, the fire

brigade and the control room will be able to effectively coordinate

fire fighting requirements with the safe shutdown of the reactor

units and is therefore acceptable.

7.5.6 Emergency Lighting

The lighting systems provided for the Browns Ferry Nuclear Plant

are described in Part X, Section 8.0 of the Plan. These systems

include (1) normal lighting, (2) standby lighting which provides

normal lighting for essential areas of the plant and is powered

from the on-site diesel generators, (3) emergency lighting which

is a separate system from the normal system and provides lighting

to the essential areas of the plant and to personnel passageways.

The level of emergency lighting is adequate for plant operation

without a fire occurring. It is continuously lit and powered by

the normal 240 volt AC power system and is automatically transferred

to the 250 volt DC station battery. However, during a postulated

fire that evolves smoke, this and other installed systems cannot be

relied upon for adequate lighting for the fire brigade.

The licensee has added portable hand lamps for emergency use

including'that of the fire brigade. These lamps are stored in the

emergency equipment cabinet in the Unit 2 control room corridor. A

total of 12 portable lamps are provided which are powered by 6 volt

batteries that have a use-life of 6 to 8 hours. This is a sufficient

number for the fire brigade and the operating crew. Procedures will

be written prior to startup to provide for quarterly voltage checks

and replacement of the batteries every nine months.

With the addition of the portable hand lamps, we conclude that

adequate emergency lighting is provided for use of the fire brigade

to perform its function.
7-14

7.5.7 Ventilation Systems

The plant ventilation systems are described by the licensee in Part

X, Section 9.0 of the Plan. The cable spreading rooms' ventilation

systems, which included modifications such as rerouting of control

and power cabling for the system and the addition of isolation dampers,

have been evaluated in Section 7.3 of this report. With the power

and control cabling rerouted outside the cable spreading room, its

ventilation system will be operable and can be manually used for

exhausting smoke from a fire in the cable spreading room when the isolation function is not required for CO2 system operation.

The Reactor Building Ventilation System performs a function for

Secondary Containment. It must isolate in order for the Standby

Gas Treatment System to perform its function of maintaining a

negative pressure in the Reactor Building to prevent exfiltration

of radioactivity in the event of a release from the Primary

Containment. Therefore, the Reactor Building Ventilation System

is designed to fail in the isolated condition. In the event that

the Reactor Building Ventilation System does fail as the result

of a fire, the Standby Gas Treatment System can be used to exhaust

smoke from the Reactor Building. The measures taken to protect against

the two divisions of cabling from being affected by a fire as described

throughout Section 7.0, ensure the availability of at least one Standby

Gas Treatment System Train.

The Control Room's normal ventilation system is designed to isolate.

on an accident signal or high radiation in the intake duct to enable

the Control Room Emergency Pressurization System to perform its function

of maintaining habitability of the Control Room. The Emergency

Pressurization System could also be used by manual initiation to

prevent smoke intrusion into the Control Room from a fire external

to the Control Room.

If the fire is in the Control Room, safe shutdown of the plant can

be conducted from the Shutdown Board Rooms. The ventilation systems

for the Shutdown Board Rooms have fire dampers in the ducts that

would close to isolate the rooms from the fire. Cooling of the

Shutdown Board Rooms is attained by self-contained air conditioning

units in each room. Thus the Main Control Room could be evacuated

and habitability of the Shutdown Board Rooms is maintained for

shutting down the plant from the back-up control boards.

We conclude that the modifications made by the licensee to provide

smoke venting are acceptable and we agree that for those systems

required to isolate for Secondary Containment and for habitability of

those control areas necessary for safe shutdown of the plantno modifications should be made for purposes of smoke control to prevent any possible

reduction in the reliability of these emergency features. For these

systems, manual operation would be allowed for smoke control if

isolation was not required at the time of the fire.
7-15

7.6 Design Changes Related to Fire Detection and Extinguishment Intended

to be Completed at the First Refueling Outage of Either Unit

7.6.1 Automatic Deluge Systems

The licensee has proposed to install additional deluge systems in

the areas where the cable trays of one division cross the trays

of the other division and where two cable tray runs of opposite

divisions approach each other directly to within a few feet and

where non-divisional cable trays are within the boundary zone of

both divisions (divisional gaps). These areas are shown on

Figures 5.2-1 through 5.2-6 of Part X of the Plan. The manual

deluge systems evaluated in subsection 7.5.1.b of this report will

be automated.

The proposed systems use a design basis of 0.3 gpm/ft 2 of vertical

projected area of the cable tray stack plus the flow required for

one 1-1/2 inch hose connection. The nozzles will be located to

provide complete coverage of the cable trays to the extent practical.

This basis was derived from the fire hazard reanalysis discussed in

a previous section of this report.

For automatic operation, the deluge systems will be zoned such that

the water discharge plus the discharge from the sprinkler systems

within the zone will be considered in the drainage requirements

for the zone area.

Each zoned deluge system will be actuated automatically by the

concurrent activation of both a smoke detector and heat detection cable

in the zone. Local manual control will also be provided in the

zone area.

Provisions will be made for (1) flow and pressure testing immediately

adjacent to the dry pipes of the fixed system, (2) flushing of

the systems, and (3) air flow tests to verify that the dry pipe portion

of the system, including the nozzles, is free of obstructions.

The licensee has stated that the valves to be used with the pre-action

systems are to meet NFPA requirements. We do not agree with the

licensee that valves (the term valves as used in this section are

considered to include their operators and initiating devices) not

tested and approved by an independent laboratory are adequate to

assure valve designs which satisfy NFPA requirements. This is based

on the recommendations of the licensee's consultant and our fire

protection consultant and fire protection personnel. It is recognized

that these pre-action valves must be designed to operate under different
7-16

flow, pressure and attached system requirements. It is standard

practice in fire protection engineering to utilize approved valves

that have been tested by an independent laboratory for the intended

function. It is also recommended by NFPA that the pre-action valves

used in automatic systems be of the approved type and also recommends

independent testing. Therefore on these bases, we will require the

licensee to utilize pre-action valves that have been tested and approved

by an independent laboratory (such as Underwriters Laboratory or

Factory Mutual) to be acceptable for their intended function.

We have reviewed the design criteria for the systems and the areas

where they will be utilized. We conclude the criteria are acceptable

pending resolution of the testing and approval of the pre-action valves

as discussed above.

7.6.2 Automatic Pre-action Sprinkler Systems

The licensee has proposed to install automatic pre-action sprinkler

systems invarious areas of the plant. These areas are shown on

Figures 5.2-7 through 5.2-15 of Part X of the Plan and include such

areas as the entire area of the reactor buildings of elevation 565.0 ft.

and the entire area of elevation 593.0 ft. except for the area on the

north wall that is covered by the zoned deluge systems, areas of

the diesel generator building and areas of the pumping station.

The design basis for the sprinkler systems is to provide a coverage

of 0.3 gpm per square foot per 5000 square feet of floor area and

that the fusible links on the sprinkler head will melt between

175°F to 225OF based on a maximum expected ceiling temperature of

150OF during plant operation without a fire. The flow rate for

the sprinklers was established based on the fire hazard reanalysis

discussed in a previous Section of this report.

The automatic actuation of any of the pre-action valves will be

from smoke detectors within the area of the system. Local manual

pushbuttons in the area for manual operation of the valve will

also be provided. Alarm provisions will be made in the control

room to indicate initiation of a pre-action system.

We have reviewed the design criteria for the pre-action sprinkler

systems and the locations where they are proposed to be installed.

Based on our review, we conclude that the design criteria are

acceptable pending resolution of the testing and approval of the

pre-action valves as discussed in Section 7.6.1.
7-17

7.6.3 Manual Dry Pipe Sprinkler System

The licensee has proposed to install a manual dry pipe sprinkler

system in each spreading room as a backup to the installed CO2

system. The proposed system will be designed to provide a fusible

link sprinkler system in the spreading rooms which will be normally

maintained as a dry pipe system. Water addition to the system will

be accomplished by manually opening two series valves, located

outside the spreading rooms, and charging the sprinkler system

with water from the high pressure fire water system. If the CO2

system has not put out the postulated fire then the fusible

links on the sprinkler heads in the vicinity of the fire will melt

and discharge fire water. Drain lines will be installed in the

spreading rooms and will be normally closed by manually operated

valves. The drains will discharge to the Turbine Building drainage

system.

The design basis for the system is to deliver 0.3 gpm of water per

square foot per 3000 square feet of floor area. The fusible link

temperature rating is 175°F to 225 0 F. The flow rate for the

sprinklers was established based on the fire hazard reanalysis

discussed in a previous Section of this report.

Our review of the proposed system indicates that as a backup

system to the CO2 system, manual operation is acceptable and the

dry pipe system is preferred to preclude potential water damage

to equipment located below the spreading rooms. We therefore

conclude that the design criteria for the system are acceptable.

7.6.4 Automatic CO2 System in Cable Spreading Rooms

The licensee proposes to automate the installed CO 2 system in each

cable spreading room. This will be accomplished by utilizing the

expanded smoke detection system and the heat detector cable system

to automatically actuate the CO2 system. The manual actuation

capability will be retained to permit fire brigade operation of

the system.

The smoke detector and heat cable systems will be designed so that

any smoke detector in coincidence with any heat detector will

actuate the system. The smoke detector will also set off an

evacuation horn and start a four minute time delay to permit

evacuation of the room if occupied.

We have reviewed the design criteria for the automation of the

spreading room CO2 system and conclude they are acceptable.

8-1

8.0 Technical Specifications

The licensee has proposed technical specifications that set forth

the limiting conditions for operation and the surveillance requirements for the fire protection systems. These specifications address

the high pressure fire water system, the CO2 system and the fire

detection system for the interim period between restart and the first

refueling of each Unit. We are currently reviewing these specifications

and will have developed final specifications prior to issuance of a

license amendment to authorize restart of Units 1 and 2.

9-1

9.0 Retest Program

9.1 Fire Protection Systems

The licensee has provided retest outlines and criteria for some

of the fire protection systems. In particular, they cover those

systems that are being modified or additions made to the original

systems. The licensee has stated that it will provide a total

pre-operational testing program including the above mentioned areas.

When this information is made available to the staff we will

complete our review and report our evaluation in a supplement

to this report prior to issuing the license amendment that authorizes

return to operation.
9-2

9.2 Program for Component and System Retests

The licensee has proposed a program for component and system retest

which is described in Part XI of the Plan. The program establishes

the administrative controls that will govern the development of test

procedures and the conduct and evaluation of the tests and describes

the tests that will be conducted prior to fuel loading (preoperational

tests) and following fuel loading (startup tests).

The objective of the preoperational retest program is to verify that

plant components and systems are in a status consistent with plant

license requirements. Testing will be conducted on all systems and

components that: (1) sustained direct or indirect fire damage,

(2) were disrupted to comply with interim license requirements,

(3) were potentially damaged by cleanup and restoration work, or (4) have

received significant modifications and maintenance. TVA's Division

of Power Production will be responsible for the overall retest program

administration, including preparation of detailed test instructions,

conduct of the tests, and review and approval of test results. TVA's

Division of Engineering Design will prepare retest scoping documents,

approve test instructions and also review and approve test results.

The licensee has not proposed to conduct preoperational load discharge

tests of the 250 vdc unit batteries. The staff will require

Tennessee Valley Authority to conduct these tests prior to fuel

loading. The staff has concluded, with the exception noted above,

that the preoperational retest program proposed by Tennessee Valley

Authority will retest systems and components that were disrupted

or potentially disrupted and verify that such systems will function

in accordance with design.

Although both Browns Ferry Units 1 and 2 have previously completed

their initial startup test programs, the licensee has proposed to

repeat selective startup tests following fuel loading. The staff

is currently reviewing this program and will complete the review

and resolve any outstanding items with the licensee prior to

issuing the license amendment that authorizes return to operation.
10-1

10.0 Conclusions

In summary, the fire damage at the Browns Ferry Units 1 and 2 has

been removed, and the facility has been restored and modified.

This has included substantial additional protection against adverse

affects of fire. The circuity has been modified to assure adequate

divisional isolation. The cabling will be thoroughly coated in all

vital safety areas with Flamemastic. This fire retardant coating

will assure that the combustible materials of the cabling do not

become sources of fire propagation. The substantially enrhanced

fire fighting and fire detection capabilities along with the

increased effectiveness of the fire brigade will assure that any fire

that may occur will be extinguished promptly, well before any

damage can occur to vital systems of the facility.

We have indicated in the foregoing sections some items that must be

resolved prior to authorizing return to operation. We will issue a

supplement to this report to present our evaluation of the resolution

of those items prior to authorizing such return to operation. We

have concluded, subject to their satisfactory resolution, that the

restoration of Browns Ferry Nuclear Plant Units 1 and 2 including

the modifications, as described in TVA's Plan for Evaluation, Repair,

and Return to Service of Browns Ferry Units 1 and 2 as a Result

of the March 22, 1975 Fire, dated April 13, 1975 and Revisions

thereto up to and including Revision 37, is acceptable and that

there is reasonable assurance that the health and safety of the

public will not be endangered by operation of the Facility as

restored and modified.

Attachment 1

Page 1

ATTACHMENT 1

CHRONOLOGY

BROWNS FERRY NUCLEAR POWER STATION, UNITS 1 AND 2

RESTORATION RELATED CORRESPONDENCE

April 1, 1975 TVA letter re fire which occurred in cable trays on

March 22, 1975, in Unit 1 spreading room and reactor

building

April 13, 1975

April 15, 1975

April 17, 1975

April 22, 1975

April 23, 1975

April 23, 1975

April 23, 1975

April 30, 1975

May 2, 1975

May 9, 1975

May 13, 1975

TVA letter

and Return

(March 22,

transmitting "Plan for Evaluation, Repair,

to Service of Browns Ferry Units 1 and 2

1975 Fire)."

TVA letter requesting amendment to licenses re the

fire which damaged numerous control cables

Letter to TVA relating to restoration of fire-affected

features, Units 1 and 2

TVA letter transmitting listing entitled, "Material to

be Provided Formally for Nuclear Regulatory Commission

Review of the March 22, 1975 Browns Ferry Nuclear

Plant Units 1 and 2 Fire."

TVA letter transmitting Revision 1 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA letter transmitting report "Physical Damage to

Electrical Cables and Raceways Involved in the

Browns Ferry Nuclear Plant Fire on March 22, 1975"

TVA letter transmitting a notebook of photos of fire

affected zone

Summary of April 24, 1975 meeting re Safety Analysis for

Removal of Fire-Damaged Components

TVA letter transmitting Revision 2 to "Plan for

Evaluation, Repair, and Return to Service"

Letter to TVA transmitting Amendment No. 9 to DPR-33

and Amendment No. 6 to DPR-52 to ensure safe condition

until restoration of facility

TVA letter transmitting Revision 3 to"Plan for Evaluation

Repair, and Return to Service ... "
Attachment 1

Page 2

-2 -

May 15, 1975

May 25, 1975

May 25, 1975

May 29, 1975

June 2, 1975

June 5, 1975

June 6, 1975

June 9, 1975

June 9, 1975

June 11, 1975

June 13, 1975

June 20, 1975

June 24, 1975

July 9, 1975

July 16, 1975

TVA letter transmitting Revision 4 to "Plan for

Evaluation, Repair, and Return to Service..."

TVA letter transmitting Revision 5 to "Plan for

Evaluation, Repair, and Return to Service..."

TVA letter transmitting Revision 6 to "Plan for

Evaluation, Repair and Return to Service..."

TVA letter transmitting Revision 7 to "Plan for

Evaluation, Repair and Return to Service..."

Letter to TVA transmitting comments on "Plan for

Evaluation, Repair, and Return to Service..."

TVA letter transmitting Revision 8 to "Plan for

Evaluation, Repair, and Return to Service ...1

TVA letter transmitting Revision 9 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting Revision 10 to "Plan for

Evaluation, Repair, and Return to Service ... 1

TVA letter advising that they will revise their procedui

as necessary to comply with the requirements indicated

TVA letter furnishing information regarding Quality

Assurance Program for the restoration on Browns Ferry

Units 1 and 2

Letter to TVA transmitting Amendment No. 10 to DPR-33

and Amendment No. 7 to DPR-52 authorization to commence

restoration

TVA letter transmitting Revision 11 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting Revision 12 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting Revision 13 to "Plan for

Evaluation, Repair, and Return to Service ... "1

Summary of July 1. 1975 meeting re design criteria for

restoration and electrical splicing
Attachment 1

Page 3

- 3 -

July 21, 1975

July 22, 1975

July 25, 1975

August 13, 1975

August 15, 1975

August 18, 1975

August 18, 1975

August 20, 1975

August 21, 1975

August 25, 1975

August 25, 1975

August 26, 1975

August 26, 1975

August 27, 1975

August 28, 1975

Summary of July 11, 1975 meeting re administrative

controls

TVA letter transmitting Revision 14 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA letter transmitting Revision 15 to "Plan for

Evaluation, Repair, and Return to Service ... "

Letter to TVA requesting additional information in

order to review "Plan for Evaluation, Repair, and

Return to Service ... "

TVA letter transmitting Revision 16 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA responses to NRC questions of August 13, 1975

TVA letter transmitting Revision 17 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA letter transmitting Revision 18 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA letter justifying total independence of

redundant systems to the point that a fire could

burn indefinitely without any reliance on

fire-fighting activities

Summary of August 19, 1975 meeting re questions

on restoration plan

TVA letter relating to removal of fire-affected

equipment listing items critical to the schedule

for startup

TVA letter transmitting Revision 1 to Design

Criteria BFN-50-D703

TVA letter relating to their August 25, 1975

letter and Part VIII to "Plan for Evaluation,

Repair, and Return to Service ...

TVA letter transmitting Revision 19 to "Plan for

Evaluation, Repair, and Return to Service ...

Letter to TVA re general items of restoration

work requiring approval
Attachment 1

Page 4

-4 -

August 29, 1975

September 2, 1975

September 2, 1975

September 9, 1975

September 15, 1975

September 15, 1975

September 17, 1975

September 23, 1975

September 26, 1975

September 30, 1975

October 2, 1975

October 10, 1975

October 10, 1975

October 28, 1975

October 31, 1975

TVA letter requesting approval of restoration

activities and revising the Plan

TVA letter transmitting Revision 20 to "Plan for

Evaluation, Repair, and Return to Service..

Letter to TVA transmitting Amendment No. 14

to DPR-33 and Amendment No. 11 to DPR-52 re

restoration work and design modifications proposed

in the Plan

TVA letter transmitting Revision 21 to "Plan for

Evaluation, Repair and Return to Service .. 1

TVA letter transmitting Revision 22 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA letter furnishing additional commitments

relating to "Plan for Evaluation, Repair, and

Return to Service ... "I

Summary of August 28, 1975 meeting re TVA plans

for restoration of BFNP

TVA letter transmitting Revision 23 to "Plan for

Evaluation, Repair, and Return to Service ..

TVA letter transmitting Amendment 62 to FSAR

relating to quality assurance

Letter to TVA re transmitting Prenotice re

authorization of operation of BFNP upon completion

of work required to restore plant

TVA letter transmitting Revision 24 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting Revision 25 to "Plan for

Evaluation, Repair, and Return to Service .

Summary of October 1, 1975 meeting re electrical

penetration seal and fire stop design

TVA letter transmitting Revision 26 to "Plan for

Evaluation, Repair, and Return to Service ... 1

Letter to TVA transmitting Amendment 15 to DPR-33

and Amendment 12 to DPR-52 re surveillance

requirements for diesel generators
Attachment 1

Page 5

- 5-

November 7, 1975

November 14, 1975

November 17, 1975

November 21, 1975

November 26, 1975

December 1, 1975

December 5, 1975

December 8, 1975

December 8, 1975

December 10, 1975

December 15, 1975

December 16, 1975

December 1 9, 1975

December 31, 1975

January 7, 1976

TVA letter transmitting Revision 27 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting Revision 28 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting "Procedures for External

Flame Test Using a Full Scale Mockup"

TVA letter transmitting Revision 29 to "Plan for

Evaluation, Repair, and Return to Service ...

Summary of November 12 and 13, 1975 meeting,

re resolution of outstanding items for fire

protection

Letter to TVA re discussions held on November

meetings re appointed coordinator, and access to

expertise in fire protection

TVA letter naming J. R. Calhoun the responsibility

of overall restoration coordinator

TVA letter advising of future design changes

in order to meet the objectives of prior TVA

commitments

TVA letter transmitting Revision 30 to "Plan for

Evaluation, Repair, and Return to Service ... 1

TVA letter transmitting Revision 31 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting Revision 32 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA letter requesting Amendment relating to

overall Restoration Coordinator

Letter to WVA transmitting Amendment No. 18 to DPR-33

and Amendment No. 15 to DPR-52 re -revision to the

Interim Technical Specifications to remove

requirements for cooling and ventilation equipment

TI/A letter transmitting Revision 33 to "Plan for

Evaluation, Repair, and Return to Service ... "1

TVA letter transmitting Special Filing Instructions

for Part X, Section A of Recovery Plan
Attachment 1

Page 6

- 6-

January 9, 1976

January 13, 1976

January 13, 1976

January 15, 1976

January 20, 1976

January 21, 1976

February 3, 1976

February 6, 1976

TVA letter transmitting Revision 34 to "Plan for

Evaluation, Repair, and Return to Service

TVA letter advising it is unnecessary to perform

a Type A primary containment leak rate test

Summary of December 3, 1975 meeting, re program

for test of equipment prior to return to operation

TVA letter transmitting data from fire tests

performed on electrical cable wall penetration

seal/fire stop designs for use in restoration

and in Unit 3

TVA letter transmitting Revision 35 to "Plan for

Evaluation, Repair, and Return to Service ... "

w/report "Metallurgical Examination of Piping

and Piping Components Which Were Exposed to

Residue from Plant Fire"

TVA letter requesting Amendment to add Tech

Specs to cover fire protection system

TVA letter transmitting Revision 36 to "Plan for

Evaluation, Repair, and Return to Service ... "

TVA letter transmitting Revision 37 to "Plan for

Evaluation, Repair, and Return to Service ... "I
Attachment 2

Page 1

ATTACHMENT 2

ROLE OF CONSULTANTS

Consultants have been utilized by the licensee and NRC in the investigation

of the March 22, 1975 fire, review of methods used to assess the damage

and cleanup of the fire effects and consultation on the restoration of

the plant to operational status particularly with relation to fire prevention,

protection and administrative procedures.

During the NRC investigation of the events and damage due to the fire,

two consultants were used by NRC's Office of Inspection and Enforcement

(I&E) to assist them in their investigation. These consultants provided

reports that are included in the July 25, 1975 inspection report, which

has been included in the Browns Ferry Nuclear Plant Docket File. These

consultant reports covered preliminary material testing for fire potential

that were installed in the plant and observations by a fire protection

consultant of the fire protection program that existed prior to the fire.

The licensee also obtained the services of two consultants to assist and

advise them on the plant cleanup methods and procedures and to advise

on the chloride effects on structures, equipment and piping in the plant.

These consultant reports are included in the licensee's "Plan for Evaluation,

Repair, and Return to Service of Browns Ferry Units 1 and 2 (March 22

1975 Fire)."

The licensee also obtained the services of a fire protection consultant

firm to assist them in developing fire prevention controls and procedures

and to advise them on the fire protection measures being proposed by the

licensee to NRC and to assist in the design modifications that are being

made to the Plant.
Attachment 2

Page 2

-2

NRC has also obtained the services of a fire protection engineering

consultant to advise it on the methods, procedures and modifications

being made by TVA as presented in the Plan.

Both the licensee's and NRC's fire protection consultants have provided

comments and recommendations that are either contained in the Plan submitted

by the licensee or the public docket record for the Browns Ferry Nuclear

Plant. Summary reports will be submitted by both consultants and copies

will be included in the Docket File prior to NRC authorizing return to

operation.

There are two areas raised by TVA's consultant where we do not yet have

sufficient information to fully evaluate the Plan in regard to fire detection

and protection. These are set forth below:

1. Based on the recommendations of the licensee's and NRC's fire

protection consultant and NRC's fire protection engineer, we believe

that the design and installation of the smoke and heat cable detection

systems should be completely reviewed by the licensee to show how they

conform with the requirements of Class 1 systems, including the

supervisory requirements. These requirements are set forth in NFPA

No. 70, 71, 72 series (A, B, C, D and E). The review should also

include those portions of the systems that were installed prior to

the March 22, 1975 fire and are now being utilized with the new systems

to function in critical areas of the plant required for safe shutdown.

The smoke detector being utilized should be shown to be compatible

with the stated requirements. If the licensee takes exception to any
Attachment 2

Page 3

-3-

of the requirements they should be stated and justification provided

NRC. Also, if the systems design and installation does not meet the

requirements for Class 1 systems these areas should also be identified

and proposals made by the licensee as to the resolution of each area.

Other recommendations made by the licensee's consultant, regarding

the design of the detection systems, should be completely addressed

by the licensee. These are recommendations 47 and 50 in Enclosure

1 of the document filed by the licensee which we received on

January 26, 1976.

2. Based on our review of the licensee's consultant's recommendations

and the licensee's responses to their recommendations, it is not

clear what the differences are in the area of hydraulic calculations

for the fire water system. In order for NRC to complete its review

of this area, we require the licensee to provide a listing of the

specific differences and their justification for the differences

being acceptable. NRC will review the hydraulic calculations after

the differences stated above are provided by the licensee.
I
Attachment 3

Page 1

ATTACHMENT 3

NRC COMMENTS ON NELPIA RECOMMENDATIONS

This comparison has been prepared on the basis of the Nuclear Energy

Liability and Property Insurance Association recommendations contained

in their report on the fire that occurred at the Browns Ferry Nuclear

Plant. As such, it must be recognized that the NRC comments are

specific to the Browns Ferry Nuclear Plant and do not necessarily apply

to other operating plants or plants under construction.

A. Cable Spreading Room (CSR) (Cable Spreading Room and similar rooms)

NELPIARECOMMENDATIONS

1. A CSR for each unit should be provided. Each spreading room

should be cut off and arranged totally independent of other CSR's

by a fire barrier wall of 3-hours' fire resistance rating.

2. At least one 3-foot wide, 8-foot high aisle should be provided

the length and width of the CSR to insure fire fighting access.

Class A fire doors should be installed at each entrance (at least

two) into the room.

NRC C(OM•MENTS

NRC has reviewed the above recommendations and does not consider them

necessary for Browns Ferry. TVA has proposed to (a) automate the

existing CO2 system in each cable spreading room, (b) coat all exposed

cable surfaces in the CSR's with a fire retartant coating (Flamemastic

71A) to retard and localize any postulated cable fire in the rooms,

(c) install a manual dry-pipe sprinkler system and floor drains in the

CSR's that will serve as a backup to the CO 2 system, (d) install
Attachment 3

Page 2

-2-

additional smoke detectors and heat cable detectors in the CSR's to

increase the capability to promptly detect a fire. We believe that

these changes, in addition to those administrative changes proposed

to reduce the probability of a fire, provide sufficient protection

in this area. Moreover, the arrangement of the Browns Ferry CSR is

such that the cost and downtime to accomplish such modifications

would not be justified in view of the limited additional benefit

that would result.

NELPHIA RECOMMENDATION

3. A standard installation of open-head, water spray sprinklers

controlled by an automatic deluge valve and products-of-combustion

actuated detectors should be provided in each CSR. The deluge

valve should be located outside the room and connected to the

station's annunciator system.

NRC COMMENTS

TVA has proposed to automate the existing CO2 system in each CSR, and

coat all exposed cable surfaces with a fire retardant coating. In

addition, instead of a pre-action deluge system as recommended by

NELPIA TVA proposes to install a manually operated dry-pipe sprinkler

system in each CSR. The manual valves will be located outside the

CSR. The proposed additional smoke and heat detection systems for

the CSR's will provide annunciation in the control room and at the

local stations. NRC finds this to be an acceptable alternate to the

NELPIA recommendation.
Attachment 3

Page 3

-3-

NELPIA RECOMMENDATION

4. One-inch hose connections, equipped with 75 ft of 1 1/2" woven

jacket lined fire hose and adjustable spray nozzle should be

provided in the CSR and located at approximately 100 ft intervals.

NRC COMMENTS

NRC agrees with the intent of this recommendation and TVA will install

three hose stations in the control building equipped with neoprene

lined hose and electrically safe spray nozzles. These hoses will be

capable of reaching any door of either CSR. Critical equipment safety

has been considered and provisions made to protect this equipment from

water.

NELPIA RECOMMENDATION

5. The concrete floor should be pitched to drain the sprinkler and

hose steam discharge to a suitable drainage facility.

NRC COMMENTS

NRC agrees. TVA proposes to install floor drains in the CSR's that

will drain sprinkler or hose water to the Turbine Building sump.

NELPIA RECOMMENDATION

6. Approved smoke and heat venting of the CSR utilizing a powered

mechanical exhaust system actuated by products of combustion

detectors should be provided.
Attachment 3

Page 4

-4-

NRC COMMENTS

The existing design is based on utilization of a CO2 system as the

primary means for fire extinguishment in the CSR. Accordingly, it

is necessary to provide for isolation of the CSR from the ventilation

system in the event a fire is detected. To assure isolation of a CSR

for operation of the CO2 system, modifications have been made which

include the addition of dampers between CSR's and the rerouting of the

ventilation system control and power cabling outside of the CSR.

These changes will provide for increased reliability for smoke and

heat venting of a CSR by manual means when CO2 operation is not required.

B. Cable Construction

NELPIA RECOMMENDATIONS

1. A re-evaluation of current cable testing requirements should be

made to establish the pass-fail criteria for flame propagation of

"real-life" cable tray systems.

2. As a minimum requirement today only those cable constructions that

will pass the current IEEE 383 flame test should be used.

NOTE: This does not infer that cables passing this test do not

require fire protection nor that certification of cables passing

more realistic tests are not essential for tomorrow's cables.

3. Whenever practical, cables that do not liberate copious quantities

of corrosive gases should be used particularly in strategic

relatively inaccessible and highly susceptible areas.
Attachment 3

Page 5

NRC COMMENTS

NRC is accepting replacement cabling purchased to the standards

existing at the time of construction. A change in cable materials is

not considered to be necessary, by NRC, in view of measures being taken

to reduce the probability of fire through application of Flamemastic

on all exposed surfaces of cables, administrative changes, and fire

detection and extinguishment systems in areas of cabling where a

fire in one division could affect redundant safety equipment.

C. Cable Tray Protection

NELPIA RECOMMENDATIONS

1. Cable tray systems should be protected by automatic, zoned,

open-head, water spray sprinkler systems arranged to discharge

directly onto the cables in the trays.

2. An approved fast-acting products-of-combustion type detection

system should be provided to actuate the deluge system having

sectional control.

3. Adequate floor drains and curbs should be provided to safely

remove discharged sprinkler water. Drainage water should be

monitored for radioactive materials before being released to the

environment. Curbs should be provided around all floor penetrations.

4. Approved noncombustible fire stop constructions should be

located in each cable tray and spaced at maximum intervals of

10 ft. horizontally and 10 ft. vertically. NOTE: Cable derating

should be given consideration when installing fire stops.
Attachment 3

Page 6

-6-

5. Wherever practical, isolate, shield, relocate water damageable

equipment.

NRC COMMENTS

NRC agrees in principle. Outside of the CSR, which is covered under

Item B, all of the above are being provided. In lieu of intermittent

fire stops TVA is providing continuous fire stop by use of a fire

retardant coating.

D. Indoor Hose Connections

NELPIA RECOMMENDATIONS

1. Fire protection equipment including hose, nozzles, standpipe

valves, and hydrants should have compatible threads with existing

equipment and the local fire department.

2. Combination spray-straight steam nozzles should be provided on

each hose connection to effectively combat Class A fires normally

inaccessible, e.g., cable tray fires.

3. Standpipe risers should be sealed on each floor to prevent smoke

and corrosive gases from penetrating into areas normally unexposed

to the effects of fire.

NRC COMMENTS

The recommendation D.1 has been corrected by TVA by providing for

new installation, compatible threads with the local fire department.

Adaptors are provided at all other stations to provide for thread

compatibility. Recommendation D.2 has been considered and TVA has

provided electrically safe spray nozzles for the hose stations.

Recommendation D.3 is covered under Recommendation K.
Attachment 3

Page 7

-7-

E. Smoke and Heat Removal

NELPIA RECOMMENDATIONS

1. Approved smoke and heat venting facilities independent of the

station's normal ventilation system should be provided throughout areas having a combustible occupancy. Each system should

be actuated by products-of-combustion detectors and arranged

to contain the release of radioactive materials.

2. The mechanical exhaust system should be powered from electrical

feeders run outside the fire area. If inside wiring is necessary,

mineral-insulated metal sheathed cable should be used.

3. Additional preventive measures outlined in the "International

Guidelines for the Fire Protection of Nuclear Power Plants"

should be implemented. Specifically Sections 6.1 (Extraction

of Smoke and Heat) and 6.2 (Preventing Corrosion) are applicable.

NRC COMMENTS

The cable spreading rooms' ventilation systems, which included

modifications such as rerouting of control and power cabling for the

system and the addition of isolation dampers, have been evaluated in

Section 7.3 of this report. With the power and control cabling rerouted

outside the cable spreading room, its ventilation system will be

operable and can be manually used for exhausting smoke from a fire

in the cable spreading room when the isolation function is not

required for CO2 system operation.
Attachment 3

Page 8

-8-

The Reactor Building Ventilation System performs a function for

Secondary Containment. It must isolate in order for the Standby

Gas Treatment System to perform its function of maintaining a

negative pressure on the Reactor Building to prevent exfiltration

of radioactivity in the event of a release from the Primary

Containment. Therefore, the Reactor Building Ventilation System

is designed to fail in the isolated condition. In the event that

the Reactor Building Ventilation System does fail as the result of

a fire, the Standby Gas Treatment System can be used to exhaust smoke

from the Reactor Building. The measures taken to protect against the

two divisions of cabling from being affected by a fire as

described throughout Section 7.0, ensure the availability of at least

one Standby Gas Treatment System Train.

The Control Room's normal ventilation system is designed to isolate

on an accident signal or high radiation in the intake duct to enable

the Control Room Emergency Pressurization System to perform its function

of maintaining habitability of the Control Room. The Emergency

Pressurization System could also be used by manual initiation to

prevent smoke intrusion into the Control Room from a fire external

to the Control Room.

If the fire is in the Control Room, safe shutdown of the plant can

be conducted from the Shutdown Board Rooms. The ventilation systems
Attachment 3

Page 9

-9-

for the Shutdown Board Rooms have fire dampers in the ducts that would

close to isolate the rooms from the fire. Cooling of the Shutdown

Board Rooms is attained by self-contained air conditioning units in

each room. Thus the Main Control Room could be evacuated and

habitability of the Shutdown Board Rooms is maintained for shutting

down the plant from the back-up control boards.

We conclude that the modifications made by the licensee to provide

smoke venting are acceptable and we agree that for those systems

required to isolate for Secondary Containment and for habitability of

those control areas necessary for safe shutdown of the plant no modifications should be made for purposes of smoke control to prevent any

possible reduction in the reliability of these emergency features.

For these systems, manual operation would be allowed for smoke control

if isolation was not required at the time of the fire.

F. Cable Penetration

NELPIA RECOMMENDATIONS

1. All wall and floor openings through which electrical cables or

conduits penetrate should be protected against the passage of

flame and smoke by devices and constructions approved or listed

by recognized testing laboratories.
Attachment 3

Page 10

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NRC COMMENTS

NRC agrees with the above protection requirements. For the applications

required for the TVA plant as-built design, no known approved or listed

devices or constructions exist. TVA has conducted extensive testing

and developed a new cable penetration fire stop design that utilizes

fire resistant inorganic materials, fire retardant coating and a silcone

RTV compound for sealing the penetrations. The wall and floor

openings will utilize this design where a pressure seal is required.

Other penetrations will utilize a design which includes fire resistant

board on each side of the penetration with the void being filled with

an inorganic fiber. The outside of the fire stop, including the cabling,

will be coated with a fire retardant coating.

NELPIA RECOMMENDATIONS

2. Temporary wall and floor openings should be sufficiently sealed

with a noncombustible material at the end of each workday to

insure the fire integrity of the wall or floor.

3. Open flames should never be used to check the installation,

gas tightness and integrity of penetration seals. Whenever

protected openings are examined, fire extinguishers should be

immediately available to those checking for openings on both sides

of the wall.

NRC COMMENTS

NRC agrees. These items have been corrected by revised plant

procedures.
Attachment 3

Page 11

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G. Self-Contained Breathing Apparatus

NELPIA RECOMMENDATION

1. Self-contained breathing apparatus approved by the United States

Bureau of Mines and described in NFPA No. 19B should be provided

for all fire fighting and control room personnel. Preferably,

their service or operating life should be one hour.

2. On-site reserve air supply should be available and arranged to

expediently replenish the supply of air in each unit so that the

designed service life is available.

NRC COMMENTS

NRC agrees; however, the service or operating life of one hour does

not appear practical to achieve. TVA has therefore increased the

available air breathing units as well as standby air cylinders. In

addition, the recharging capacity has been doubled and an air

compressor has been added.

H. Physical Independence of Redundant Circuits

NELPIA RECOMMENDATIONS

1. All redundant Class IE circuits and the equipment served by

these circuits should be separated from the primary Class IE

circuits by a minimum three-hour fire wall. This will require

that a redundant cable spreading room be constructed.

2. Mineral-insulated metal sheathed cable or equivalent fire

resistant cables should be used in one of the two Class IE

electrical circuits.
Attachment 3

Page 12

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NRC COMMENTS

NRC does not believe these actions are necessary in light of other

actions taken. (Please see responses given for Items A.1, A.2, and B.1,

B.2 and B.3).

I. Cardox Total Flooding System

NELPIA RECOMMENDATIONS

1. The ventilation system in the CSR should be arranged to shut down

whenever the Cardox system is discharged.

2. The Cardox system should be rearranged to operate automatically

upon actuation of the ionization detection system. NOTE: A

one-minute delay should be incorporated into the system to

allow workers ample time to leave.

3. A written procedure and permit system should be adopted that

would require employees to obtain written permission to impair

fire protection equipment.

4. An acceptance test of the fire protected system, including a

complete discharge, should be conducted and witnessed by the

installer and authority having jurisdiction.

5. An investigation into the compatability of the ionization

detectors with the products-of-combustion generated by the

burning cable should be made to insure that the detectors will,

in fact, operate during the incipient stages of the fire.
Attachment 3

Page 13

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NRC COMMENTS

NRC agrees. All items have been accomplished. NOTE: In regard to

Item 2, a four minute delay has been incorporated. In regard to Item 4,

the testing was accomplished prior to initial startup of Units 1 and

2.

J. Control Room (CR)

NELPIA RECOMMENDATIONS

1. All floor openings between the CSR and CR should be sealed airtight

with a material that will insure the fire resistance integrity

of the floor. Only penetration seals listed by Underwriter's

Laboratories or approved by the Factory Mutuals should be

considered. Cellular concrete, and inorganic assemblies as

described in the "International Guidelines" may also be considered.

NRC COMMENTS

NRC agrees with intent to seal floor openings between CSR and CR

and as stated in our comments on recommendation F.1, TVA has developed

and tested a design that will accomplish this function.

NELPIA RECOMMENDATION

2. Self-contained breathing apparatus approved by the United States

Bureau of Mines should be located in the CR to insure an orderly

station shutdown and to minimize breathing hazards to personnel.

The supply should be sufficient for the number of operators and

the time it takes to effect a safe shutdown.
Attachment 3

Page 14

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NRC COMMENT

NRC agrees. TVA has emergency equipment cabinets that contain storage

for approved self-contained breathing apparatus for the control room

personnel and the fire brigade. Additional storage of available units

is located in another area of the plant.

K. Stairwells, Vertical Opening, Mechanical Penetrations

NELPIA RECOMMENDATION

1. All stairwells, elevators, chutes and other vertical openings

should be enclosed in approved masonry towers with airtight,

automatic closing Class B fire doors at each opening into the

building.

2. All unprotected vertical openings between floors (hoistways

steam pipes, etc.) should be sealed airtight.

NRC COMMENT

The Reactor Building is a complex structure with requirements for

established ventilation patterns to achieve proper distribution and

control of potential radioactivity. There is no practical way to

a sh this recommendation for isolation of floors etc. The

intent of the recommendation, however, i.e., to minimize the spread

of the products of combustion, has been achieved by minimizing the

contained combustible material including Flamemastic coating of exposed

cables. In addition, improvements have been made in fire related

administrative procedures and automatic detection and extinguishment

systems have been added.
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