ML18033A777

had

been

stopped

and the maintenance

request

(MR) was written which included the

successful

comple-

tion of the

SI to satisfy

the

post-maintenance

testing

PMT

requirements.

The radiation monitor was

removed,

repaired,

and

was reinstalled at 1:00 p.m.,

on February

25,

1989.

Technical

Specification table 3.2.A, note

G requires

the Unit

1

Reactor

Building to

be isolated

and the

Standby

Gas

Treatment

(SBGT)

system

to

be started

when the instrument

channel

which

provides

the

"Reactor

Building Ventilation

High

Radiation

Reactor

Zone"

function is inoperable.

This radiation monitor

performs that function for channel

"A".

During the

performance

of the SI, the

SBGT system trains are initiated and the Unit

1

reactor

zone is isolated until the completion of the test.

At

the time the test deficiency

was identified and the radiation

monitor was

taken

out of service,

these

TS

requirements

were

satisfied

as

a direct result of the performance of the SI.

Once the radiation monitor was reinstalled,

the SI was continued

until 6:05 p.m.,

on February

25,

1989,

when the

was

stopped

due

to lack of coverage

on

the night shift.

At that time,

operators

reset

the

SBGT system trains

and the Unit

1 reactor

zone isolation

as well

as other

equipment

actuated

during the

SI.

The

was started

again

on

February

26,

1989,

at 8:00

a.m.,

and the

TS requirements

were met at 8: 17 a.m.; but the SI

was

stopped at 8:50 a.m.

because

a higher priority SI was

needed

for Unit 2.

Once

again

the

SBGT system

was

shut

down and the

Unit

1 Reactor

Building isolation was reset

when the SI was

stopped.

On

February

28,

1989,

licensee

personnel

identified

that the

PMT (completion of the SI) had not been

completed

and

the radiation monitor.was declared

inoperable.

The compensatory

actions

requi red

by the

TS were reinitiated at

10:20 a.m.,

on

February

28,

1989.

Containment,

filtration,

and controlled release

of postulated

radioactive

releases

are specific functions of secondary

contain-

ment

and

Standby

Gas Treatment

(SBGT).

The ventilation exhaust

radiation monitors provide automatic isolation/acutation

signals

which are

required

for the

secondary

containment

and

SBGT to

perform their functions.

'I

Per

the

TS definition of Secondary

Containment Integrity,

is required to be operable

and:

All the unit reactor

building ventilation

system

penetrations

required to be closed during accident conditions are either:

1.

Capable

of

being

closed

by

an

operable

reactor

building

ventilation

system

automatic

isolation

system or

2.

Closed

by

a least

one

reactor

building ventilation

system

automatic

isolation

valve deactivated

in the

isolated position.

Therefore,

the

operability

of

and

automatic

secondary

containment

isolation

require

the operability

of the

exhaust

radiation monitors.

TS 3.2.A states

that the instrumentation

required for primary

containment

integrity is

given

in

Table

3.2.A

and

states

that the reactor vessel,

reactor building, main

steam lines,

and

are

also

included.

Table

3.2.A

requires

1

radiation

monitor

channel

for each

of the

two train

systems.

If the

required

number of channels

is

not met for the reactor

zone

exhaust

radiation monitors

then

the

reactor

building

must

be

isolated

and

SBGT started.

1n

summary only one radiation monitor was operable

on the unit

1

r eactor

zone ventilation exhaust for a p'eriod of several

days

following corrective

maintenance

because

the post

maintenance

testing

had not been

performed.

The required

number of operable

radiation

monitor

channels

was

not

met

and

the

compensatory

actions

for this

condition

were

not

maintained

while

the

radiation monitor was inoperable.

4

e

The

failure

to maintain

secondary

containment

and

the

Technical Specification 4.6.B,

system

trains

in operation

during

the

time period that

the

Unit 1,

channel

"A" reactor

zone

exhaust radiation monitor was

inoperable is considered

a violation of Technical Specification 3.2.A, 3.7.B,

and 3.7.C (Violation 259,

200, 296/89-11-01).

The

licensee

identified

and reported this violation to the

NRC

as

documented

in LER-89006,

issued

March 30,

1989.

An

NRC notice

of .violation (NOV) will be issued,

rather than classifying it as

"licensee

identified" with no

NOV since

the corrective action

did not identify the

steps

being

taken

to

ensure

that

post

maintenance

testing will be completed promptly.

On March 7,

1989,

the licensee identified that

a reactor coolant

sample

was not performed

as required

by

TS after

the

Unit

3

continuous

conductivity monitor (3-CIT-43-11)

was

removed

from

service

on March 6,

1989 at 9: 15 a.m.

"Coolant Chemistry," requires

that

a

sample of reactor

coolant

be analyzed

every

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for conductivity and chloride

ion content

when the continuous conductivity monitor is inoper-

able.

On March 7,

1989,

at 6: 15 a.m.,

a

sample

was

taken

for

the

normal surveillance

frequency

and at 7:30 a.m.,

a chemistry

lab technician

reported that the monitor was out of service

and

initiated the

8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

sample

frequency.

The

monitor

was

removed

from

service

for calibration

and

troubleshooting

on March 6,

1989 per

MR 877517.

From

a review

of the

completed

MR package,

the

NRC inspector

noted that

a

TS

time limit was not entered

on the

MR and the working instruc-

tions indicated that

a work impact evaluation

was not required.

The

NRC inspector

also

noted that the

SOS

and Unit

1 operator

logs did "not list the monitor as

being

inoperable

on

March 6,

1989.

However,

the

"Work

Log

Sheet"

provided

with the

MR

package. stated

that the operator

was notified of the monitor's

removal

and return to service.

The failure to perform

a reactor coolant

sample analysis within

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the

removal of the Unit 3 continuous

conduc-

tivity monitor

from service

is

considered

a violation of

Technical Specification 4.6.B. 1.c,

(Violation 296/89-11-05).

Although this violation

was identified

by the

licensee,

the

inspector. did not believe that the licensee's

correction action

to prevent recurrence

was adequate

and

a 'NOV will be issued.

I

From reviews of the control

room logs,

completed MR'ackages,

LREDs,

and associated

incident critiques, it appears

that the errors

were

the result of problems in the planning,

review,

and implementation of

maintenance

activities

and the affected operability requirements

for

safety-related

equipment.

Also, insufficient review,

or

lack of

review,

of the

MR packages

resulted

in neither of the

components

being declared

inoperable.

by operations

personnel.

The

NRC inspec-

tors are concerned

that these

examples

occurred

due to similar causes

and

may be indicative of a generic

problem.

b.

Diesel Generator

Walkdown

The

NRC inspector

walked

down

System

82,

Units

1/2

and

3 Diesel

Generators

during this reporting

per iod using

the following plant

drawings:-

0-15E500-3

Unit

1 and 2,

Key Diagram of Standby

Auxiliary Power System

3-15E500-3

Unit 3,

Key Diagram at Normal

and Standby

Auxiliary Power System

1-47E859-1

Unit

1 and 0,

Flow Diagram Emergency

Equipment Cooling Water

Unit 0,

Flow Diagram Fuel Oil System

c.

Failure to Follow Special

Operating Instruction

On April 5,

1989,

the licensee

discovered that Shutdown

Board

B was

not lined up as required

by a special

operating instruction

(memo

B22

890404

014).

Due to various plant activities, additional

480 volt

loads

were shifted to the

shutdown board, i.e. the

2A 480V shutdown

board

and the

A Diesel Auxiliary Board.

In order to preclude

the

possibility

of

an

overload

condition,

the

special

requirements

in

this special

operating instruction stated that the

1C and

2C

RHR pump

breakers

.were

to

be

racked

out

and disabled.

However, it was

discovered

that the

2C

RHR

pump breaker

had

not

been

racked

out.

This item was

evaluated

and determined

by the licensee

as not being

reportable.

The inspector

considers

the action taken

by the licensee

to

be appropriate.

A violation for failure to follow a procedure is

not being cited because

the criteria specified in Section

V.A. of the

Enforcement Policy were satisfied.

This is identified

as

Non-cited Violation

(NCV) 260/89-11-04

for

which

no

response

is required.

The

inspectors will continue

to

monitor licensee activities in this area.

0-47E840-3

The

NRC inspector

noted that during this reporting period,

various

D/Gs for all units

were placed in and out of service to support the

Division I, Division II and Division III outages.

No deficiencies

were noted

as to valve lineups, control availability, and switchboard

availability.

The

components

in all eight

DG rooms

and the valves

directly outside the

rooms were adequately

labeled

and identified to

support operation of the system.

Two

violations

and

one

non-cited

violation

were

identified

in

the

Operational

Safety Verification area.

Surveillance

Observation

(61726)

The

inspectors

observed/reviewed

the

surveillance

instructions

(SI)

procedures

discussed

below.

The inspections

consisted

of a review of the

SIs for technical

adequacy

and

conformance

to TS, verification of test

instrument calibration, observation

of the conduct of the test,

confirma-

tion of proper

removal

from service

and return to service

of the

system,

and

a review of the test data.

The inspectors

also verified that limiting

conditions for operation

were

met, testing

was

accomplished

by qualified

personnel,

and the SIs were completed at the required frequency.

On

February

28,

1989,

the

licensee

discovered

that

an expired

Immediate

Temporary

Change

(ITC)-5 was still in place for O-SI-4.7.B.3.C,

"Standby

Gas

Treatment

System

Train Operability Test."

The

SI demonstrates

the

operability of the

SBGT system trains

by manually starting

each train and

verifying that the fan starts,

the relative humidity heaters

energize,

and

dampers align as required.

The ITC was

issued

to allow the verification

of

the

acceptance

criteria

with

the

trains

already

in operation.

Specifically, it allowed the performer to skip the steps for starting the

trains if they were already

running.

At the time the

ITC was written, the

SBGT system trains were running to satisfy other

TS requirements.

The

ITC was given

an expiration date of February

13,

1989,

and should

have

been

removed

on that date

in accordance

with

SDSP

2. 11,

"Implementation

and

Change

of Site

Procedures

and

Instructions."

The

NRC

inspector

reviewed

copies of the

SI which were

performed daily from February

21,

1989

through

February

24,

1989,

for the

"B" and

"C"

system trains

with the expired

ITC still in place.

In each

performance,

the

SBGT system

trains were not running

and the changes

allowed by the

ITC were not used.

Operability of the trains

was verified in accordance

with the approved

steps.

SDSP 2. 11 requires that

ITCs be

removed

from the affected

procedures

upon

expiration.

If the

system

trains

had

been

running during the

performances

on February

21,

1989 through February

24,

1989, the

ITC steps

could

have

been

used

and

TS surveillance

requirements

could

have

been

missed

due to the performance

of an unapproved

pr'ocedure.

Administrative

controls

are

established

to

require

activities

to

be

performed

in

accordance

with requirements

and strict

implementation

of the controls

ensure that the requirements

are met.

This concern

was discussed

with the

licensee.

No violations or deviations

were identified in the Surveillance

Observa-

tion area.

Maintenance

Observation

(62703)

Plant

maintenance

activities

of

selected

safety-related

systems

and

components

were observed/reviewed

to ascertain

that they were conducted

in

accordance

with requirements.

The following items were considered

during

this review:

the limiting conditions for operations

were

met; activities

were

accomplished

using

approved

procedures;

functional

testing

and/or

calibrations

were

performed prior to returning

components

or systems

to

service;

quality

control

records

were

maintained;

activities

were

accomplished

by qualified

personnel;

parts

and

materials

used

were

properly certified;

proper tagout

clearance

procedures

were

adhered

to;

and

radiological

controls

were

implemented

as

required.

Maintenance

requests

were reviewed to determine

the status of outstanding

jobs and to

assure

that priority was assigned

to safety-related

equipment

maintenance

which might affect plant

safety.

The inspectors

observed/reviewed

the

,following maintenance activities during this report period:

On March 5,

1989, while performing post maintenance

testing

(PMT) on

the

"1/2 A" diesel

generator

(DG) following scheduled

Division I

outage

work,

the

DG automatically

started.

This

occurred

when

maintenance

personnel

misinterpreted

a step

in the instruction

and

tho'ught that the

DG monthly operability test (O-SI-4.9.a. l.a.(A)) was

to

be

performed

as part of the

PMT.

Operations

began

an electrical

lineup checklist in accordance

with the SI and due to the electrical

lineup already

in place

per the

PMT, the

DG received

a start signal

and started

as designed.

Operations

shut

down the

DG and realigned

the manipulated

components

to the positions

they were in prior to

beginning the SI.

The

PMT procedure

was clarified and

completed

as

intended.

The

DG was

tagged

out

and

had not been declared

operable

at the time of the event.

On March 10,

1989,

the inspector

observed

WR 911557 which was written

to

troubleshoot

the

cause

of

the

Unit

3

channel

"A" trip

(half-scram actuation)

which occurred

on March 7,

1989, while the bus

was connected

to its alternate

power source (transformer).

Following

the trip, control

room annunciations

indicated the cause

to be

a loss

of power to the

"A" RPS

bus

~

However, field examination

found the

bus to

be energized

and

none of the circuit protectors (for under-

frequency,

undervoltage,

and overvoltage)

were tripped.

The licensee

considered

the

cause

to

be

a fluctuation in the

bus voltage

which

temporarily

made

up the circuit protector logic and then returned to

normal.

The circuit protectors

contain

two time delay relays in the

circuit to trip the bus

and the

system. engineer

stated that there

had

been

problems with the circuit protectors

in the past.

During the

troubleshooting,

the

licensee tried to duplicate perturbations

that

may have

been

seen

on the bus

power supply by starting

a control

bay

chiller.

The results indicated that the voltage drop was observable,

but it did not decrease

to the circuit protector trip setpoints.

The

system

engineer

stated

that the operation

of the circuit protectors

and future problems would continue to be reviewed until

a root cause

and

permanent

correction

could

be established.

The

NRC inspector

noted significant involvement by operations,

system

engineering,

and

maintenance

in

planning

and

conducting

the

troubleshooting

activities.

No discrepancies

were identified.

The

resident

inspectors

followed

licensee

maintenance

activities

associated

with problems

with the Unit

2

RHR

Loop II injection

isolation valve,

2-FCV-74-67.

The valve which is. a

24 inch

motor

operated

gate

valve with

a

SMB4T Limitorque actuator

had failed

on

March

28,

1989.

During post modification testing,

the

valve

was

backseated-

and

the

thermal

overload

device

prevented

any further

operation.

The valve

had

apparently

become

backseated

during

the

process

of checking for proper direction of rotation after personnel

had failed to place

the valve position

near

enough

to mid position

prior

to attempting

motor

operation

of the valve.

MR 869976

was

written to disassemble,

inspect

and repair the valve

and actuator.

The valve was disassembled

and

no apparent

damage

was detected.

As

a

precaution,

an

ultrasonic

test

of

the

valve

stem

and

visual

inspection

of the

stem

threads

were

performed

with

no

apparent

defects

found.

After repairs

under

MR 869976 were performed,

MOVATs

testing

on 2-FCV-74-67

was

performed.

During the

MOVATs testing

on

April 5, the limitorque operator

again failed when it was

unable

to

unseat

the valve disc.

MR 916006

was written to again disassemble,

inspect

and

repair

the

valve

and

actuator.

The

actuator

was

partially disassembled

and

cleaned,

but valve disassembly

was not

required.

The

l,icensee

determined

that

the failure

was

due

to

a

compressed

spring pack in the actuator.

The

NRC inspector

determined

from discussions

with licensee

personnel

that evidence of spring pack

degradation

was

indicated

in the thrust

signature

during

MOVATs

testing prior to the failure.

The spring

pack

was replaced,

valve

stem

checked for evidence

of bending

and

damage,

valve stroked to

determine

any damage,

actuator

reassembled

and applicable portions of

MOVATs testing reperformed.

The

licensee

performed

an

engineering

evaluation

to determine if

overstress

or other

damage

occurred

to the

valve during

the

two

events.

During the

evaluation,

TVA contacted

the

Limitorque

and

Crane-Aloyce

companies

to obtain their recommendations

and

comments.

According to the

vendors

292,000

lbs is the

maximum possible thrust

that could

have

been

applied

by the actuator

during

a locked motor

condition.

The

weakest

member,

the

stem threads,

are

expected

to

begin yielding at 384,627 lbs.

No actual thrust higher than

194,000

lbs

was measured

during

MOVATs testing of the valve.

Based

on this,

the licensee

determined

that

no overstress

occurred

and

the

valve

should

be safe to operate.

The

NRC inspector

reviewed

the

work packages

associated

with

MR

869976

and

MR 916006.

The packages

contained

a sufficient amount of

detail

and

guidance

to allow acceptable

accomplishment

of the work

and

the

documentation

supported

that

adequate

post

maintenance

testing

was performed.

The

NRC inspectors

walked

down portions of Hold Order 0-89-183,

and

Caution Order 0-89-145 concerning

RHRSW pump maintenance

and reviewed

the clearance

logs to verify compliance with SDSP-14.9,

"Equipment

Clearance

Procedure."

The

review verified

that

the

clearances

isolated

the

affected

portions

of the

systems

being

tagged;

the

documented

components

were tagged

and in the correct positions;

and

any applicable

TS limiting conditions for operation

were satisfied.

No discrepancies

were identified.

~

e

No violations or deviations

were identified in the Maintenance

Observation

area.

5. 'esign

(37700)

The

NRC inspectors

expressed

concern

about the significant number of

instances

of failure to neet the single failure design criteria that have

been recently identified:

During

a licensee restart test

program review of the 250 Volt Battery

Boards

and the

480 Volt AC Shutdown

Boards that occurred in January

1987,

a system engineer

determined that the loss of 250 Volt Battery

Board

Number

1 would cause

loss of 480 Volt Shutdown

Boards

lA 5 1B

due to load shed signals.

The result of this condition was the loss

of the Division II core spray logic.

This concern

was documented

by

the

NRC as IFI 259,

260, 296/88-04-04.

A design deficiency

was identified during the restart test

program

review by the licensee

when both signal trains for the

standby

gas

treatment

system

logic were found wired

up to the

same

relay which

closes all four dampers

located

in the equipment

bay.

This concern

was documented

by the

NRC as IFI 259,

260, 296/88-05-06

'uring

the recent

review by the

NRC of concerns

associated

with the

seismic affects of the vitrified clay piping in the

RCW system, it

was noted that both redundant'afety

related air conditioning units

associated

with the Unit 2 4KV Shutdown

Board

Rooms would be rendered

inoperable

due to

a single fai lure of the

common

EECW di scharge

flow

path resulting

in the failure of both divisions of 480 Volt Reactor

MOV Boards.

This design

was

the result

of

a

recent

modification

associated

with

ECN

P0956.

Additional detail

is

included

in

NRC

Inspection

Report 259,

260, 296/89-10.

The licensee

recently identified an unanalyzed

condition

on March 22,

1989,

where

the

A,

B,

and

C Diesel

Generators

could

be

overloaded

during the first few minutes of an accident.

This could possibly

occur due to the failure of a single lockout relay which prevents

the

three

motor driven fire

pumps

from starting

during

an

accident

condition including

a loss of coolant accident or a loss of offsite

power.

The single relay is

shared

among

the

three

pump starting

logic circuits.

The fire pumps

are

designed

'to automatically start

on

low fire protection

water pressure

with only one

pump starting

initially.with a second

pump and perhaps

a third pump starting if the

10

first pump is unable to provide adequate

system pressure.

Each fire

pump receives

electrical

power

from

a separate

diesel/shutdown

buss

and

a possibility

exists

that

the

three

affected

diesels

could

successionally

fail

on overload

because

the

respective

fire

pump

motors were not blocked for the first 10 minutes of diesel

loading

as

required.

The fire.

pump

demand

could

be

due

to

actuation

of

non-qualified fire detectors

during

an

accident

condition.

This

issue is documented

in

CARR BFP 890219.

The

NRC inspectors

met with members of licensee

management

and requested

that the licensee

described

the

guidance

relative to detecting

potential

single failure problems that they give their personnel

that perform design

and design

change

reviews,

review of test results,

and in review events.

The licensee

has yet to provide

a satisfactory response'o

this concern.

This potential failure to provide adequate

single fai lure criteria during

the design

process will be

documented

as

an

Unresolved

Item (259,

260,

" 296/89-11-02)

pending further review of the single failure critdria used

by the licensee.

Resolution of IFIs

259,

260, 296/88-04-04

and 259,

260,

296/88-05-06 will be included in the resolution of this single

URI which

must

be resolved prior to restart of Unit 2.

No violations

or deviations

were identified in the

Design Deficiencies

area.

6.

Fire Prevention/Protection

- (42051)

An

NRC inspector toured the Units 1, 2,

and 3, reactor buildings, control

bay,

and turbine building to observe

the fire prevention

and protection

activities.

The

inspector

verified the

following conditions

to

be

in

effect:

wood scaffolding was marked

as being treated with flame retardant

there

was

no

unnecessary

accumulation

of combustible

forms,

form

lumber, shoring,

or scaffolding

fire extinguishers

and fire hoses

were

located at designated

places

at each elevation

access

to suppression

devices

was not restricted

by outage materials

or equipment

suppression

devices

indicated current inspection.

0

The

NRC inspector identified one concern in that there

was

a large

amount

of material

contained

in yellow plastic

bags

stored

within several

contamination

areas

located in the Unit 3 Reactor Building.

Since

some of

the yellow plastic

bags

appeared

to possibly contain combustible material,

the

NRC inspector identified this concern

to licensee

management.

After

evaluating

the condition, the licensee

stated that all of the material in

question

was

stored

in approved

locations

and that all

bags

contained

mi rror type insulations or other noncombustible material.

~ 1

e

No violations

or deviations

were

identified in the

Fire Prevention/

Protection

area.

Cable Deterioration

(62703)

During the performance

of a functional test of the Unit 2

IRM A channel

on

March

14,

1989,

a short circuit occurred

resulting

in the

power . supply

fuse blowing for that drawer.

The licensee

investigated

the cause of the

short

circuit

and

determined

that

various

cables

-in

the

Nuclear

Instrumentation

System were deteriorated

in that the rubber insulation

had

become brittle and

cracked with portions of the conductor

exposed.

This

problem is documented

on

CAQRs BFP890290,

BFP890291,

and

BFP890292.

V

The licensee

reported this problem to the resident

inspectors

on March

17.,

1989.

Based

on preliminary information, the

problem

was believed to

be

limited to General'lectric

supplied

GENIE

SJO

SI - 53115

power

cables

with black

Nitrile-PVC jacket.

The

conductor

insulation

is

styrene

butadiene

(Buna-S)

rubber insulation which was

known to have

been

used in

the

source,

intermediate,

and average

power range instrumentation

systems

at Browns Ferry.

The failure of the insulation material

appears

to

be embrittlement

and

cracking of the individual conductor's

rubber insulation after prolonged

exposed to air in the area

where the outer jacket was stripped

away during

termination

of

the

cables.

The

ozone

concentration

in

the

normal

atmosphere

causes

the material

to deteriorate.

Neither the cable

outer

jacket or conductor insulation

covered

by the jacket is expected

to

be

affected.

The licensee

is in the

process

of replacing

the

NI cables with properly

qualified cable.

The Division I NI cables

have

been

replaced

with the

Division II work remaining

to

be

completed.

Licensee

management

has

stated

that although

GE

has notified the licensee that based

on

a sample

review of local'anels

there is probably

minimal

use

of thi s

type of

cable,

TVA has

requested

that

GE perform

a complete

review of local panels

for application.

Since

there

was

a possibility that the

same

type of cable

may have

been

used

on

other

Nuclear

Steam

Supply

System

(NSSS)

components

GE

was

requested

to

manually

search

their

records

and

determine.

any

other

applications

for this type cable.

Based

on preliminary information from

General Electric,

17 other

BWRs built between

the late

1960s

and

1971 also

have this type cable

and

no prior problems with this type cable

have

been

identified.

GE replied via

GE Electrical

Design

Engineering

Memos dated

March 24,

and

March 27,

1989, that their review of control

room drawings

for Browns Ferry Unit 2 was

complete

and that this material

was

used

as

power

cable

for other

systems

in addition

to NIs.

Process

control

instrumentation

(GE-MAC), Area

Radiation

Monitors

(ARMs),

and

various

control

room recorders

were identified as also

being affected.

GE also

stated

that

a

RICSIL communication

(type of

immediate

action

SIL)

was

12

being issued to

BWR owners to provide formal notification and

recommenda-

tions.

GE

made

a preliminary assessment

of available information associated

with

the problem and determined that it was not an immediate threat to safety.

This

assessment

is partial ly

due

to

the= fact that

the

associated

components

would fail due to loss of power if a electrical

short occurred,

resulting in any required

RPS or

ESF actuation

occurr'ing, i.e.

component

failure would result in fai lure in a nonconservative

manner.

GE has obtained

a .sample of the cable for further analysis to identify any

additional

reasons

for the deterioration

and will'be providing information

to

INPO on the issue.

The

NRC inspectors will follow the licensee's

progress

in this area during

future inspections.

Specifically the

NRC inspectors

are

concerned

about

the

proper identification

and

replacement

of defective

cables

in all

applications 'of this type cable in control

room and local panels

through-

out the plant.

Additionally the

NRC inspector will need to review any

licensee

and/or vendor generic evaluation that is performed.

This item is

identified as Inspector

Followup Item 259,

260, 296/89-11-03;

Deteriorated

GE Cables.

This item must

be resolved prior to restart of each respective

Unit.

No violations-=or deviations

were identified in the

Cable Deterioration

area.

8.

Reportable

Occurrences

(90712,

92700)

, The following licensee

events

reports

(LERs) were reviewed to determine:

adequacy

of event description,

verification of compliance with technical

specifications

and

regulatory

requirements,

corrective

action

taken,

existence

of potential

generic

problems,

reporting requirements

satisfied,

and the relative

safety significance .of each

event.

Additional in-plant

reviews

and 'discussions

with plant

personnel,

as

appropriate,

were

conducted.

0

(OPEN)

LER

296/89-03:

Unplanned

Engineered

Safety

Features

Actuations

Caused

By Voltage Transient

on Electrical

Distribution

'System.

On March 7,

1989,

Unit 3 received

an unplanned

ESF actuation

due to

voltage fluctuations in the alternate

power supply to the

RPS circuit

protectors

3Cl and

3C2, which caused

a momentary loss of power to

Bus 3B.

The normal

3B

RPS bus is supplied

by the

RPS motor generator

set

number

3B

and

the alternate

supply is

from

a unit preferred

!

0

e

13

regulatory transformer.

The

NRC inspector

reviewed the

LER as it may

affect Unit 2

power supplies

which are

the

same electrically

as

Unit 3.

The following drawings were reviewed:

45W641-4,

Powerhouse

Unit 2,

Wiring

Diagrams,

Power

System

. Schematic

Diagram SH-4.

2-45E641-2,

Powerhouse

Unit 2,

Wiring Diagram

Instrumentation

and

Control

Power System,

Schematic

Diagram.

The

NRC inspector

noted that the alternate

power for the Unit 2

A

and

B is

a direct feed from the secondary

of a step

down transformer

480 to 120/240

Y, Unit Preferred

Regulatory Transformer

TVP-2, and is

consequently

susceptible

to voltage fluctuations

induced

by the

480

volt primary side

shared

loads

and the

240/120 volt secondary

side

shared

loads.

The

normal

supply to the

Bus

A and

Bus

B is

provided

by

a

480 to

120 volt motor-generator

set

equipped with a

flywheel and does

not have

a shared

load system

on the output of the

generator.

This

system

is

not readily susceptible

to voltage

and

load'fluctuations

on the

480 volt shared

power feed to the motor due

to

the

flywheel

and

the

electrical

isolation

provided

by

a

motor-generator

system.

The

NRC

inspector

also

noted

that

by

transferring

the Unit 3

Bus

3A to the alternate

source

in order

to perform

a

PM on March 1,

1989,

and by not performing the

PM in

a

timely manner," i.e.

when the event occurred approximately

seven

days

after the transfer,

the

PM still had

not

been

performed, this left

that

Bus

much

more susceptible

to electrical fluctuations.

The

licensee

stated

in the Unit 3

LER that administrative

steps

would be

taken

to

minimize

the

amount

of time all three

unit