Text

Insp Repts 50-250/89-52 & 50-251/89-52 on 891202-22. Violations Noted.Major Areas Inspected:Monthly Surveillance & Maint Observations,Operational Safety,Fitness for Duty Training & Plant Events
	ML17347B544
Person / Time
Site:	Turkey Point
Issue date:	01/22/1990
From:	Butcher R, Mcelhinney T, Orlenjak R, Schnebli G; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
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ML17347B542	List: ML17347B541; ML17347B543; ML17347B544;
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	50-250-89-52, 50-251-89-52, NUDOCS 9002010246
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UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

101 MARIETTASTREET, N.W.

ATL AN TA, G Eo R G IA 30323

lf

Report Nos.:

50-250/89-52

and 50-251/89-52

Licensee:

Florida Power

and Light Company

9250 West Flagler Street

Miami,

FL

33102

Docket Nos.:

50-250

and 50-251

Facility Name:

Turkey Point

3 and

4

License Nos.:

DPR-31

and

DPR-41

Inspection

Conducted: 'ece

ber 2,

1989 through

December

22,

Inspectors-:- ~

.

r-c- ~ c

R.

C.

cher,

nior

R

i ent Inspector

Mc-iZ.

T.

F.

cElhinne,

Resi

ent Inspector

~r

l c.i'~

G. A.

bli, Resid

t Inspector

Approved by:

R.

.

rlenjak,

ection

Ch

f

Division of Reactor Projects

1989

Date Signed

/8~

'Pd

D te Signed

SUMMARY

Scope:

This routine resident

inspector

inspection

entailed direct inspection

at the

site

in the

areas

of monthly surveillance

observations,

monthly maintenance

observations,

operational

safety,

fitness for duty training and plant events.

Results:

There

was

one violation with two examples,

two URIs**, four IFIs, and

two NCVs,

identified.

Also,

the

residents

expressed

a

concern

regarding

two

recent

events

that resulted

from operators'nattention

to detail

and the fai lure to-

maintain communications

during

EDG runs which was

a contributing factor in one

of the events.

e

" Unresolved

Items

are

matters

about

which more information is required to

determine

whether they are acceptable

or may involve violations or deviations.

. 900i22

pgp

QQIOCK

Q

50-250,251/89-52-01,

IFI.

Excessive

acceptable

leak rate

through fire

protection-PIV.

(Paragraph

2)

50-250,251/89-52-02,

IFI.

Modify procedures

to ensure

personnel

notified

of site evacuation

alarm

and evacuation

has occurred.

(Paragraph

7)

50-250,251/89-52-03,

IFI.

Verification of tygon

tubing indication for

BAST level indicator channel

check.

(Paragraph

7)

50-250,251/89-52-04,

IFI.

Initiation of

PM requirements

to prevent lint

accumulation

in laundry room.

(Paragraph

9)

50-250,251/89-52-05,

Violation,

two

examples.

Failure

to

follow

procedures

resulting in inadvertent

release

of liquid waste

from the

B. MT

(paragraph

9) and failure to follow procedure resulting in the

B

EDG speed

droop not being adjusted during

a Surveillance Test.

(Paragraph

5)

50-250,251/89-52-06,

NCV.

Failure to maintain

RCP seal injection throttle

valve closed resulting

in

an inadvertent 'increase

in

RCS level while in

Mode 5.

(Paragraph

3)

50-250,251/89-52-07,

NCV.

Failure

to incorporate

required testing

into

plant procedures

for ICW isolation valves

4-4882

and

4-4883 to

TPCW heat

exchangers.

.(Paragraph

3)

50-250,251/89-52-08,

URI

~

Followup

on investigation

of

NCR 86-421 being

closed without required actions

being completed.

(Paragraph

4)

50-250,251/89-52-09,

URI.

Followup

on

licensee's

corrective

actions

regarding failure to establish

TS required fire watch in time following

Auxiliary Building evacuation.

(Paragraph

9)

REPORT DETAILS

Persons

Contacted

Licensee

Employees

T. Abbati el 1 o,

equal ity Assurance,

Supervi sor

J.

W. Anderson, guality Assurance

Supervisor

J. Arias, Technical Assistant to Plant Manager

"L. W. Bladow, equality Assurance

Superintendent

R.

M. Brown, Health Physics

Supervisor

J.

E. Cross,

Plant Manager - Nuclear

R. J. Earl, guality Control Supervisor

T. A. Finn, Assistant Operations

Superintendent

S.

M. Franzone,

Lead Engineer

S.

T. Hale,

Engineering Project Supervisor

K. N. Harris, Vice President

G. Heisterman,

Electrical Assistant

Superintendent

RE J. Gianfrencesco,

Assistant

Maintenance

Superintendent

~V. A. Kaminskas,

Technical

Department

Supervisor

J.

AD Labarraque,

Senior Technical Advisor

G. Marsh,

Reactor

Engineering

Supervisor

R. G. Mende, Operations

Supervisor

"L. W. Pearce,

Operations

Superintendent

D. Powell, Regulatory

and Compliance Supervisor

K. Remington,

System

Performance

Supervisor

G.

M. Smith, Service

Manager - Nuclear

F.

H. Southworth,

Assi stant to Site Vice President

R.

N. Steinke,

Chemistry Supervisor

J.

CD Strong,

Mechanical

Department

Supervisor

"G.

S. Warriner, guality Control, Supervisor

M. B. Wayland,

Maintenance

Superintendent

J.

D. Webb, Assistant Superintendent,

Planning

and Scheduling

Other

licensee

employees

contacted

included

construction

craftsman,

engineers,

technicians,

operators,

mechanics,

and electricians.

Attended exit interview on December '22,

1989.

Note:

An Alphabetical Tabulation of acronyms

used in this report is

listed in paragraph

11.

Followup on Items of Noncompliance

(92702)

A review

was

conducted

of the following noncompliances

to assure

that

corrective actions

were adequately

implemented

and resulted

in conformance

with regulatory

requirements.

'erification

of corrective

action

was

achieved

through record reviews,

observation

and discussions

with licensee

personnel.

Licensee

correspondence

was evaluated

to ensure

that the

e

'responses

were timely and that corrective actions

were implemented within

. the time periods specified in the reply.

(Closed) Deviation 50-250,251/88-36-02.

Failure to have

a test

program

as

required

by the

FSAR to demonstrate

conformance

to design

and

system

readiness

requirements

for the Fire Protection

System.

On

October

29,

1988,

an event

occurred

where

a portion of the fire main could not

be

isolated

due to leaking PIVs.

This condition resulted

in both fire pumps

being

declared

OOS.

The

licensee

in

response

to this

event

issued

Surveillance

Procedure

OSP-016.30,

Fire Main Post

Indicator

Valve

Leak

Test

and

System

Flush,

dated

June

29,

1989.

The inspector

reviewed

the

procedure

and found the licensee's

method for determining

PIV leakage

to

be acceptable

for those

leak rates

less

than

50

gpm.

The basis for the

maximum allowable

leakage

rates

for the

specific fire main isolation

valves

identified

in

Enclosure

1 of the

procedure,

were

established

through

the

use of

a computer

model.

The

computer

model

evaluates

the

various sprinkler system

demands,

hose

stream flow, and pressure

require-

ments,

flow paths,

fire

pump

flow and

pressure

requirements

and

then

calculates

the

remaining

system

flow which could

be lost due to system

leakage

without

the fire

suppression

(water)

system

being

declared

inoperable.

The inspector

reviewed the computer

model calculation results

for PIV-23 and PIV-33,

and

found the results

to

be reasonable,

based

on

fire suppression

system

water flow demands

in the event that

a fire were

to occur concurrent

wi,th

a pipe break requiring

PIV-23 or PIV-33 to

be

isolated.

The inspector

in reviewing,

O-OSP-016.38,

found the five gallon

bucket

and stop watch leak rate measuring

method f'r leak rates

less

than

50

gpm to

be feasible.

However, for leak rates

in excess

of 50

gpm this

, method

could not

be easily

implemented.

For example,

PIV-33

has

a leak

rate of 590

gpm, this would require

118 five gallon 'buckets of water to be

collected

in

a

one

minute

time frame.

In response

to this concern

the

.

licensee

initiated

a request

for -procedure

review

and

provided

a draft

revision of procedure

O-OSP-016.30

to

the

inspector

for review.

The

licensee

draft'rocedure

includes

provision for calculating

flows in

excess

of 50

gpm using data gained

by Pitot Tube

Flow measurements.

The

inspector

agrees

that

the

l.icensee

through

the

implementation

of this

revised draft test

procedure

can

demonstrate

system

design

conformance.

This item is closed.

However, the inspector is still concerned

that with

established

acceptable

leak rates

in excess

of the Jockey

Pump Capacity

and with

a pipe break in the

underground fire main, both fire pumps

may

still have to

be placed

OOS

in order

to facilitate repairs.

This is

identified as IFI 50-250,251/89-52-01,

exce'ssive

design acceptable

leakage

rates

through

PIVs may require total fire suppression

system to be placed

.

OOS to facilitate repairs

in the event of fire main pipe break.

(Closed) Violation 50-250,251/88-21-01.

Concerning

the Failure to Follow

Procedure

to Place Transfer Switches

on Channel

IV for Steam Generator

"B"

Testing.

In order to prevent recurrence,

the licensee

modified procedures

of this nature

during which incorrect

channel

selection

could place

the

unit in a transient to have the channel

selections

independently verified

by

a qualified operator.

In addition,

identification labeling

of the

subject

equipment

was modified to reflect the protection

channel

numbers.

This item and

LER 50-251/88-10,

(discussed

in paragraph

4) are closed.

(Closed)

Violation

50-250,251/89-12-03.

Failure

to fol low

procedur es

resulting

in

valve

201B

being

open

and

draining

RCS

water

to

the

containment

sump.

The licensee,

in response

to this violation, performed

a review of procedure

AP-0103.4,

In Plant

Equipment

Clearance

Orders,

to

determine

procedure

adequacy with respect

to vent and drain

hose install-

ation and vent/drain valve manipulations.

In order to prevent recurrence,

the

licensee

revised

procedures

3/4-0P-41.8,

Filling and

Venting

The

Reactor Coolant System

and 3/4-OSP-041. 1, Reactor

Coolant

System

Leak Rate

Calculation.

Procedure

3/4-OSP-041. 1,

was

revised

to incorporate

the

requirement

to visually

inspect

RCS

vent

and

drain valves'ith

hoses

connected

to assure

no

leakage

is occurring.

In addition,

the inspector

verified that procedure

3/4-0P-41.8

was

revised

to require that

whenever

RCS 'filling and

venting

operations

are

occurring that

one

containment

sump

pump

and

one

channel

of

sump

level

indication

be

operable.

The

inspector

found the licensee's

actions

in response

to this violation, to be

satisfactory.

This item is closed.

(Closed)

Violation

50-250,251/89-18-01.

Fai lure

to

fol low

procedure

resulting

in

the

inadvertent

actuation

of Train

A Safeguards.

The

licensee,

in response

to this violation, placed

warning

signs

inside

the

safeguard

racks of both units

and at the

power supply breakers

to these

racks.

The warning signs inside the safeguard

racks state that installing

fuses

FU4 or

FU3

may result

in

an

SI being

generated.

This sign also

references

that

procedure

3/4-0NOP-049,

Re-energizing

Safeguards

Racks

After Loss of Single

Power

Supply,

should

be utilized when installing

these

fuses.

The sign at the

power

supply

breakers

identifies

to

the

operator that closing these

breakers

can also result in an SI signal.

In

addition,

the licensee

revised the standard

clearance, for this activity in

the plant clearance

network to require that procedure

3/4-ONOP-049 is to

be

used

when releasing

clearances

associated

with the

safeguards

racks.

The inspector

found the licensee's

corrective actions associated

with this

event to be satisfactory.

This item is closed.

Followup

on

Inspector

Followup

Item(s),

Inspection

and

Enforcement

Information Notice( s),

IE Bulletin(s) (information only), IE Circular(s),

and

NRC Request(s)

(92701).

(Closed)

URI 50-250,251/88-11-01.

Evaluate

Licensee's

Method of Testing

Check

Valves to

Meet

the

Requirements

of

ASME

Code

Section

XI.

The

inspector

reviewed

OP-0209. 1,

Valve Exercising

Procedure,

dated

May 17,

1989,

to determine, that

LHSI pressure

boundary isolation

check valves,

4-876A,

B, and

C, are included in their

Valve Exercising

Program.

These

valves

are classified

as

valves

which cannot

be exercised

during plant

.operations

and

have

been

included in Appendix

B to OP-0209 '

as valve

which are

required to

be exercised full open

whenever

a unit is in Cold

Shutdown.

The inspector

also verified that the licensee

had developed

a

method for determining

the flow split between

v'alves

4-876B

and

C,

and

incorporated

the

method

into the

procedure.

The licensee's

corrective

actions

and

code

test

requirements

associated

with

check

.valves

is

consistent with the requirements

of ASME Boiler Pressure

Code,

Section XI,

Division I,

Subsection

IWV,

1980 Edition through Winter

1981

Addenda;

IWV-3520, Tests for Check Valves.

This item is closed.

(Closed)

URI 50-250,251/89-12-02.

This item concerned

RCP seal injection

throttle valve 4-297A being

opened contrary to

GOP 4-305,

Hot Standby

to

. Cold Shutdown.

This resulted

in

an

increase

in the reactor, drain

down

water level

with Unit

4 in

Mode

5.

Investigation

revealed

that

work

needed

to

be

done

on

4-297A during the outage'herefore,

the clearance

boundary

was modified and the tag

on 4-279A was

removed to allow the work

to

be

done.

When

the

maintenance

was

completed,

the

clearance

was

released

prior to re-tagging

and reclosing

4-279A allowing water from the

CVCS to enter

the

RCS through the

RCP.

The Unit 4

RCO noted the increase

in level

and

promptly identified the

source

of inleakage.

Valve 4-279A

was closed

and tagged

as required.

The fai lure to reclose

and tag 4-279A,

as specified

by

GOP 4-305, after maintenance

was

completed

constitutes

a

violation of

TS 6.8. 1.

However,

this violation

meets

the criteria

specified in

10 CFR 2,

Appendix

C,

Section V.A,'herefore,

no notice of

violation will be

issued.

The

licensee

was

in the

process

of and is

continuing to'eview the

equipment

clearance

procedure

and training of

personnel

to prevent recurrences.'his

item is closed

and will be tracked

as

NCV 50-250,251/89-52-06.

(Cl osed)

URI

50-250,251/89-24-03.

Resoluti on

of

Document

Contr ol

Discrepancies.

During

the

same

time

frame

the

document

control

discrepancies,

noted

in

NRC

IR 89-24

were identified,

the

licensee

was

conducting

a

QA Audit in the

same

area.

The

inspector

reviewed

the

results

of

QA Audit Report

QA-O-PTN-89-988,

dated

May 23,

1989,

which

covered

the audit conducted

on May 11,

1989 thru May 22,

1989, in the area

of Document Control.

Based

on the inspectors

review,

the

scope

of the

document discrepancies

identified by

NRC IR 50-250,251/89-24

were included

in the findings of the licensee's

audit report.

The inspector

reviewed

the

response

to the audit report

and QI-6-PTN-1,

Document Control, dated

August 17,

1989,

and

AP 0190.86,

Document Control, dated October 27,

1988.

The inspector

found response

to this audit report to be satisfactory

and

the

scope

of the

QI was consistent

wi,th the

scope

of the

AP governing

document control.

This item is closed.

(Closed)

IFI

50-250,251/88-30-06.

Corrective

Actions

to

Prevent

EDG

Overload

From

ICW

Pump

Auto Start

During'ccident

Conditions.

The

licensee

implemented

PC/M 88-393,

ICW Overcurrent Trip, which modified the

control

circuits

of

the

ICW

pumps

to

defeat

the

overcurrent trip

auto-start

interlocks

between

the

pumps.

This

modification,

as

implemented,

prevents

the train "B" EDG from potentially being overloaded

during

a

DBA concurrent

with

a loss of off-site power.

The inspector

reviewed

the

PC/M

and

verified this

modification

was

completed

on

November

17,

1988. This item is closed.

(Closed)

IFI 50-250,251/88-39-02.

Resolution of Root

Cause

for Unlanded

Spade

Connection

in the

EDG Control Panel.

The. licensee,

under

PWO 5390,

ER69 relanded

a pair of wires to the

ES 200

RPM

relay and the

EZ

40

RPM

relay.

The

cause

of this condition

was attributed to the routing

and

pulling of

cable

B4B0698P

inside

the

control

panel.

During

the

installation

of this cable,

the

cable

rubbed

up against

the affected

stake-on

push type lugs,

causing

these

wire connections

to

become

loose.

The inspector

agrees this was the most probable

cause for this condition.

This item is closed.

(Closed) IFI 50-250,251/88-40-03.

Followup on Licensee's

Determination of

the

Cause

For

Leakage

Through

Seal

Table

Conduit at J7

and J12.

The

licensee

had Westinghouse

perform

a failure analysis

on the Unit 3 leaking

conduit

guide

tubes.

The

inspector

reviewed the Westinghouse

Analysis

WCAP-12146,

dated

March 13,

1989.

The analysis

concluded that the type of

cracking

observed

is

indicative

of

transgranular

stress

corrosion

cracking.

In addition,

the analysis identified that the localized pitting

on

the

shoulder

and

socket

outside

diameter.

surfaces

is indicative of

attack

from a corrodant

such

as chlorides.

The metallographic analysis of

the

fracture

face

and

the

crack tip region

revealed

chlorides

were

present.

This chloride contamination,'long

with the subject

tubes

being

wetted

from the

leaks

observed

on the

swage

lock fittings., coupled with

the

stresses

from the

RCS

service

pressure

and

from the

cold

worked

microstructure

promoted

this

transgranular

stress

corrosion

cracking

failure.

The cracks

propagated

from this outside diameter

surface

in the

shoulder

and

socket

region to the. inside diameter.

The inspector

agrees

with the results of the licensee's

analysis.

This item is closed.

(Closed)

URI 50-250,251/89-45-05.

This

item

concerned

the failure to

perform quarterly

IST for

ICW isolation valves

4-4882

and

4-4883 to the

TPCW heat

exchangers.

The

cause

of this

event

was

determined

to

be

personnel

error.

AP

0190. 15,

Plant

Changes

and Modifications

(PC/M),

Figure J,

System Acceptance

Turnover

Sheet,

required

the

PUP coordinator

to list all procedures

not required

for

system

acceptance/turnover,

but,

required revision.

This section

.was

marked

N/A by the

PUP coordinator.

The need to revise

OP 209. 1, Valve Exercising Procedure,

was identified by

PUP

on October 26,

1988'n June

27,

1989,

a Request for Procedure

Review

was initiated by

PUP group to revise

OP 209. 1.

However, during the review.

process,

comments

concerning

test

frequencies

were not

resolved

at

the

reviewer level.

This

was

not elevated

to higher levels of management.

Therefore,

the testing

was not done prior to the

due date

of August

1,

1989.

The failure to

perform

the

ASME,

Section

XI, testing

of valves

4-4882

and

4-4883,

constitutes

a violation of

TS 4.0.3.

The inspectors

determined

this violation

met

the criteria

specified

in

10 CFR 2,

Appendix C, Section

G. l. for a Licensee Identified NCV.

The basis for the

determination

was

as follows:

(a) the licensee identified the 'violation,

(b) the violation was classified

as Severity Level IV, (c) the event

was

reported

to

the

NRC via

LER 50-251/89-13,

(d)

the

licensee

initiated

corrective

" actions

which included

adding

the

valves

to

OP

0209. 1

on

October

24,

1989,

counseling

the

personnel

involved

and .initiating

a

review of procedures

requiring revision for PC/Ms to identify any other

needed

procedure

changes,

(e) the violation,was

not willful and,it could

not

have

been

prevented

by corrective actions for,a previous violation.

The inspectors. mentioned

in IR 50-250,251/89-45,

other recent

examples

of

failure to incorporate

design

requirements

into plant procedures.

The

root causes

for those

events

was inadequate

administrative controls,

where

in thi s

case,

the

administrative

controls

were

adequate

but

were

not

implemented.

Therefore,

the

corrective

actions

for

the,

would

not

have

prevented

this

occurrence.

URI

50-250,251/89-45-05

is closed

and this

item will

be

tracked

as

NCV

50-250,251/89-52-07.

(Closed)

IFI 50-250,251/89-12-01.

Possible

Improper Oil Level in the

AFW

Governor.

The

licensee,

in

response

to this

item,

revised

the

PWO

guidance

associated

with checking

the oil levels

in the

AFW governors.

The

guidance

presently

utilized provides

a definition with regard

to

normal

levels

under

standby

and

operating

conditions.

The

inspector

reviewed

PWO

69-8207,

69-7458,

and

69-7459

and verified this

revised

guidance

was being properly incorporated

in the

PWO work descriptions.'n

addition,

the

licensee

incorporated

caution

notes

into

the

AFW

surveillance

procedures

with regard to the actions to be taken if the oil

level in the governor is not within its normal level range.

The inspector

reviewed

3-OSP-075. 1,

and

4-OSP-075. 1,

both

dated

July ll,

1989,

and

verified this guidance

was properly incorporated

into these

procedures.

The inspector

found the licensee's

corrective actions

to be satisfactory.

. This item is closed.

(Closed)

IFI 50-250,251/89-18-03.

Recurrence

of Voids in the

RCS

~

The

licensee

performed

an

investigation

of

the

sudden

pressurizer

level

decreases

which occurred

onJanuary

25,

1989,

and March 9,

1989, while the

reactor

was in

Mode

5, depressurized

and vented.

The inspector

reviewed

the

results

of

the

licensee's

investigation.

This

investigation

determined that

a gradual

increase

in pressurizer

level occurred prior to

each

event

along

with

a

subsequent

pressurizer

level

decrease

of

approximately

2.4%.

The licensee

determined this condition is indicative

of dissolved

gas being released

into the reactor vessel

and forming

a void

in the reactor

head area.

The gas formation in the

head

area is caused

by.

the nitrogen cover gas

on the

VCT being entrained

by the charging flow and

being

released

when

the

flow velocity

and

pressure

decreases

in the

, reactor.

The

licensee's

investigation

also

determined

that

nitrogen

venting of the reactor

head

area

is restricted

by the resistance

of the

head vent orifice and the loop seal

which is incorporated

in the head vent

system

design

configuration.

Based

on this

investigation it

can

be

concluded that the increase

in pressurizer

level is

a result of gas void

formation

in the

reactor

head

area

and

the

sudden

drop in pressurizer

level

can

be attributed to sudden

venting of collected

gas in the reactor

head

area.

The

licensee

based

on

the

results

of their investigation

revised

procedures

3/4-0P-041.7,

Draining The Reactor Coolant System,

and

attachment

to

3/4-OSP-201. 1,

RCO Daily Logs.

The

inspector

reviewed

procedures

3/4-0P-041.7,

dated

June

29,

1989,

and

3/4-OSP-201. 1,

Attachment

7,

dated

June

29,

1989,

and

determined

the

investigation

recommendations

associated

with establishing

and maintaining

a reactor

head

vent path

had

been

incorporated.

In addition,

the

inspector

has

determined

the corrective

actions

are

satisfactory

and

reactor

vessel

level perturbations

under similar operating conditions which were present

on January

25

and

March 9,

1989,

should

be reduced

as

a result of 'these=-

actions.

This item is closed,

Onsite

Followup

and

In-Office Review of Written

Reports

of Nonroutine

Events

and

10 CFR Part 21 reviews (92700/90712/90713)

The

Licensee

Event Reports

and/or

10 CFR Part 21 Reports

discussed

below

were

reviewed

and

closed.

The

inspectors

verified

that

reporting

requirements

had

been

met,

root

cause

analysis

was performed,

corrective

actions

appeared

appropriate,

and

generic

applicability

had

been

considered.

Additionally, the inspectors

verified that the

licensee

had

reviewed

each

event,

corrective actions

were

implemented,

responsibility

for corrective

actions

not ful.ly completed

was clearly assigned,

safety

questions

had

been evaluated

and resolved,

and violations of regulations

or

TS conditions

had

been identified.

When applicable,

the criteria of

10 CFR 2, Appendix C, were applied.

(Closed)

50-250,251/P2186-03.

This deficiency

involved

damage

to

lead

wire insulation for Limitorque

MOVs

DC Motors manufactured

by Peerless-

Winsmith.

The failure mechanism

was attributed

to insulation

damage

due

to bending during field installation

and setup.

FPL experienced

a failure

of a Peerless

Winsmith

DC motor

on Unit 3

AFW Steam

Supply

MOV-3-1403,

in December

1986.

Root cause

analysis

determined

the lead wire failures

were due to mechanical

damage

during packaging,

shipping,

or installation.

The

licensee

performed

inspections

and testing

of all

the

subject

DC

motors which indicated

no damaged

lead Hires.

NCR 86-421

was written to

address

this

problem.

Engineering

resolution

required

the

plant

to

replace

MOV-3-1403 with a qualified motor and to return

any

spare

motors

to the

vendor for lead wire sleeving.

The licensee identified two spare

motors with suspect

serial

numbers.

One spare

motor was in the Electrical

(}C locker and the other spare

motor was located in the nuclear warehouse.

The

NCR was

closed

out

due to MOV-3-1403 being

replaced

and

the

spare

motors

being returned to the vendor for electrical

lead replacement.

The

~ inspectors

verified that

MOV-3-1403'as

replaced

by

reviewing

the

completed

work order

and

performing

a field walkdown.

The

inspectors

could

not verify the

spare

motors

were

returned

to the

vendor.

The

licensee

determined

that this task

was not completed contrary to the

NCR

disposition.

One spare

motor

was still in the nuclear warehouse,

without

any

gC

hold

ML17347B544

Text

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