ML17342A854

(AFW)

Train

1

Operability

Verification

and

procedure

3-0SP-075.2,

Train

2

Operability

Verification.

The

licensee

re-performed

the

surveillance

procedures

in

their

entirety

and

instructions

in the night order

log were

issued

to increase

personnel

awareness

of the necessity

to complete

surveillance

procedures

or obtain

an approved

procedure

change.

This item is closed.

(Closed) Violation 250,

251/85-30-04,

Failure to identify the root cause

of AFW pump overspeed trips.

The maintenance

department

reviewed the root

cause

section

of

procedure

O-ADM-701,

Plant

Work Order

Preparation.

Additionally, the

licensee

has

established

an

event

response

team to

review and determine root causes

to problems.

This item is closed.

(Closed)

Violation 250/85-42-01,

Two

examples

of failure to follow

procedures;

one concerning

source

range

nuclear instrumentation,

and the

other concerns

the

condensate

system.

The individuals involved received

counselling,

a reminder to follow procedures

was placed in the night order

log and the two examples

resulted in LERs which were trained

on during the

1985-1986

Cycle

V operator

requalification class.

Additionally, procedure

3/4-0P-073,

Condensate

System,

was revised to require plant clearances

for

steam generator

pump motor breakers.

This item is closed.

(Closed) Violation 250, 251/85-23-01,

Failure to meet the requirements

of

10 CFR 50 '9 in the

use

of the

spent

fuel

pool

(SFP)

cooling

system

contrary to the

FSAR.

The

licensee

has

revised

procedures

O-ADM-100,

Procedures

Preparation,

Review

and

Approval;

AP-0109. 1,

Preparation,

Revision, Approval,

and

Use of Procedures;

AP-0109.3,

On The Spot Changes

to Procedures;

and AP-0109.6,

Temporary

Procedures,

to provide

improved

guidance

to individuals

responsible

for preparing

procedure

changes.

These

revisions

included

guidance

on

conducting

FSAR

and

Technical

Specification, reviews.

A special

NRC inspection

was

conducted

(Report

250,

251/87-24)

to

evaluate

engineering

procedures

and

controls for

engineering

evaluations.

This inspection

concluded that the licensee

had

adequate

controls for conducting

safety evaluations.

Additionally, the

licensee

is

taking

long

term

corrective

actions

in

the

area

of

10 CFR 50.59

reviews

as

a result of Enforcement, Action 86-20.

This item

is closed.

(Closed) Violation 250, 251/84-09-05,

Failure to adequately

review design

changes

on safety-related

electrical

busses.

The licensee

stated that the

design

changes

were not performed

due to being classified

as

non-nuclear

safety related

design

changes.

Procedure

AP 0190. 15,

Plant

Changes

and

Modifications (PC/M), was revised to require all

PC/Ms be reviewed

by the

Plants

Nuclear Safety Committee.

This item is closed.

(Closed)

Other

251/83-39-01,

Failure

to

maintain

adequate

procedure

specifying reporting requirements.

Procedure

AP 0103. 12, Notification of

- Significant Events to NRC, dated April 14,

1987

, appears

adequate

in the

reporting requirements for reactor trips.

This item is closed.

(Closed)

Other

251/83-39-03,

Failure to implement

procedure

following a

reactor trip.

The licensee

has

implemented

new procedure

3/4-0NOP-059.3,

Nuclear

Instrumentation

Malfunction,

dated

January

23,

1987,

which

requires

the operator

to confirm shutdown

margin

when both

source

range

channels

are malfunctioned.

This item is closed.

(Closed)

Violation

251/83-39-06,

Failure

to

follow procedure

when

conducting

a reactor

startup.

Procedure

3/4-0NOP-028,

Reactor

Control

System Malfunction, contains

guidance

on determining

when

a control rod is

misaligned.

This item is closed.

(Closed)

Violation

251/83-39-08,

Failure

to

implement

off-normal

procedures

for

a failed Pressurizer

Power

Operated

Relief

Stop

Valve.

Procedure

ONOP-1208. 1, Pressurizer

Power Operated Relief System - (Reliefs

and MOV's) Malfunction, dated July 24,

1986, provides adequate

guidance

to operators

on what to do for a failed stop valve.

This item is closed.

(Closed) Violation 251/83-39-09,

Failure to process field procedure

changes.

The

licensee

has

stressed

procedure

verbatim

compliance

policy,

additionally

these

requirements

are

in AP-0103.2,

Responsibilities

of

Operators

and Shift Technicians

on Shift and Maintenance

of Operations

Logs and Records,

This item is closed.

(Closed)

Violation 251/83-39-11,

Failure

to

implement

procedures

when

conducting

a feedwater

system periodic test.

The plant management

issued

circulars

to all

plant

personnel

stressing

procedural

compliance.

Procedure

AP-0103.2

was revised to require procedural

verbatim compliance.

This item is closed.

(Closed)

Violation 251/83-39-12,

Failure of on shift operators

to take

action to investigate

problems.

The licensee

restressed

the importance of

responding to problems

and taking corrective action by the operators,

this

was placed in the night order log.

This item is closed.

(Closed)

Viol'ation 250/86-17-01,

The

cont~ol

room operator

failed to

properly

implement

the unit startup

procedures.

The Plant

Manager

Nuclear

re-issued

a letter

to all

nuclear

plant

personnel

on

the

importance

of

verbatim

procedural

compliance.

New

procedures,

3/4-GOP-301,

Hot Standby to Power Operation

and 3/4-GOP-503,

Cold Shutdown

to Hot Standby,

have

been

implemented with improved guidance.

This item

is closed.

(Closed) Violation 251/83-38-01,

Both Unit 4 Containment

spray

pumps were

inoperable

during power operations.

This

was

cause

by having the

manual

header

stop valves inadvertently closed.

The licensee

took the following

corrective actions.

The operations

management

held

meeting

to discuss

.

incident;

the valves in question

were placed

under Administrative Control

as

locked

open

valves

in procedure

O-ADN-205, Administrative Control of

Valve,

Locks and Switches;

the labels

and locks for these

valves

have

been,

color coded to help prevent wrong unit/wrong train events.

This item is

closed.

(Closed)

Violation

250,251/83-38-02,

Failure

to notify the

NRC

on

10 CFR 50.72 events.

The operations

personnel

received

instructions

on

reporting requirements.

This item is closed.

(Closed)

Violation

250,251/84-04-02,

the

licensee

failed to

provide

adequate

procedures

or

to

control

the

operations

of safety

These failures resulted

in a

breakdown

in management

control

of plant operations.

The procedural

deficiencies

noted in this violation

have

been

corrected.

Additionally,

the

licensee

- implemented

the

Performance

Enhancement

Program

(PEP)

to

improve

overall

plant

performance.

This program is on-going

and its progress is being tracked

by the

NRC.

This item is closed.

(Closed)

Violation 250/84-29-03

and

251/84-30-04,

The

PNSC

did

not

adequately

review facility operations

in that potential

safety hazards

in

the Intake Cooling Water System,

Component

Cooling Water System,

Emergency

Containment

Coolers,

1.20 Volt AC Vital Bus Inverters

and remote

shutdown

instrumentation

were not detected.

The deficiencies

not in the violation

have

been

corrected

and the procedures

have

been

revised.

The licensee

has

implemented

a program for improved operation.

This item is closed.

(Closed) Violation 250,

251/82-24-02,

Failure of personnel

to follow the

requirement

of a Radiation

Work Permit

(RWP)

on protective clothing.

The

individual involved were counselled.'dditionally,

the licensee

requires

all personnel

entering

the radiation controlled

area to read

and

sign

a

log stating that they understood

the requirements

of the

RWP.

This item

is closed.

(Closed)

Violation

250,251/86-25-01,

Failure

to

properly

implement

OP-1004.2,

Reactor

Protection

System - Periodic Testing,

and

OP-4304. 1,

EDG - Periodic

Test.

The operators

involved were

counselled

on the

importance'of procedural

adherence.

Procedure

OP-1004.2

has

been replaced

by

3/4-OSP-049. 1,

Reactor

Protection

System

Logic Test

and

Procedure

OP 4304. 1 has

been

replaced

by 0-OSP-023. 1, Diesel

Generator

Operability

Test.

The deficiencies

noted in the old procedures

have

been corrected.

This item is closed.

(Closed)

Violation 250,251/85-13-01,

Failure to implement

procedures

in

the area of contaminant exclusion, radiation work permit requirements

and

housekeeping.

Procedures

HP-3207.2,

Residual

Heat

Removal

Pump

Disassembly

and

Repair,

and

Procedure

MP-1407.7,

Reactor

Vessel

STUD

Tensioner

Operators,

have

been

revised to include contaminant

exclusion

requirements.

The

individuals

involved

in the

RWP

and

housekeeping

non-compliances

were counselled

and specific training developed to address

each of these

problems.

This item is closed.

(Closed)

Violation 250,251/85-02-1,

Diesel

Generator

exceeded

voltage

limit during full load rejection testing.

The

NRC staff reviewed this

concern

and determined that

a short duration voltage transient

was not of

concern

as

long as the

EDG voltage stabilized at or below the limiting

voltage.

The licensee

has

submitted

a

TS change

to. limit the transient

time to two seconds

following the load rejection.

This item is closed.

It should

be

noted that twelve of the

above violations that are closed-

involved failur'es to implement and/or follow procedures.

The licensee's

corrective action

in the past

has primarily

been counselling,

procedure

changes,

and

issuance

of reminder letters

concerning

the

importance of

verbatim compliance with procedures.

The

NRC continues

to

be concerned

over

the

licensee's

program

to correct this continuing

problem.

These

violations in this report

have

been

closed

since there is

no practical

reason

to further track

these

individual

examples.

The

Performance

Enhancement

Program

has

included

several

projects

specifically

aimed

toward

the

improvement

of and

compliance

with procedures

(Project

2,

Operations

Enhancement

and Project 5, Procedures).

The

NRC will continue

to closely monitor the licensee's

progress

in the implementation of these

programs.

Followup

on

Unresolved

Items

(URIs),

Inspector

Followup

Items (IFIs),

Inspection

and,Enforcement

Information Notices (IENs), IE Bulletins (IEBs)

(information only), IE Circulars (IECs),

and

NRC Requests

(92701).

A review was conducted of the following items to assure that the licensee

completed

adequate

applicability reviews,

made appropriate

distributions

and if required,

implemented

adequate

and timely corrective actions.

(Closed)

URI 250,

251/85-03-02,

Throttling of

RHR Discharge

Stop Valves,

This

item

was

changed

into

a violation in

inspection

report

250,

251/86-44.

This item is closed.

(Closed)

IFI 250,251/85-06-05.

Maintenance

attention

needed

for chronic

problems

with

the

area

radiation

monitoring

system

(ARMS).

From

January

1986 to August 1986,

four contract

18C technicians

were

employed

full time to mai'ntain

the

ARMS and process

radiation monitoring

(PRM)

Systems.

From September

1986, to the present,

two

FPL

18C technicians

have

been

assigned

to these

systems

on a full time basis.

Using the

same

technicians

to perform maintenance

on these

systems

enabled

them to become

more experienced

with the

equipment,

improved the quality of maintenance

and contributed to increased

system reliability.

The licensee

plans

to=

replace

the existing

system

in the future

on

a priority derived through

the Integrated

Schedule

process.

S

(Closed)

URI 250, 251/85-20-04,

Licensee

personnel

may not be adequately

familiar with

some technical

specifications

(TS).

This unresolved

Item

addresses

a

concern

for the failure to

comply with

TS

Surveillance

requirements

and for removal

of mechanical

snubbers

without performing

evaluations

required

by TS 3. 13.3 as identified in the following LERs:

LER 250-85"01

LER 250-85-08

LER 250"85-09

LER 250-85-11

These

LERs were previously addressed

and closed in Inspection

Report 250,

251/86-39.

Additionally,

missed

TS

Surveillance

requirements

were

addressed

as

a violation (250,251/86-39-02)

and

subsequently

closed

in

Inspection

Report 250, 251/87-10.

This item is closed.

(Closed)

IFI 250,251/85-06-06.

Develop

a

procedure

for operating

the

spent fuel pool

(SFP)

leakage detection

system.

This IFI was resolved

by

revising OP-0204.2.

Subsequently,

the daily requirement to check for SFP

leakage

was incorporated into O-OSP-201.2,

SNPO Daily Logs.

This item is

closed.

(Closed)

URI

250,

251/85-20-03.

Evaluate

the advisability of blocking

safety injection while maintaining

hot standby

conditions.

The licensee

has

modified 3-GOP-305,

Hot Standby to Cold Shutdown

Procedure,

to more

clearly define the conditions which must be satisfied to place the safety

injection

block switch in the

block position.

These

conditions

were

verified to be in conformance with current

TS requirements.

This item is

closed.

(Closed)

IFI

250,251/85-24-08.

Improve

procedural

guidance

for

containment evaluation

alarm and high flux at shutdown

alarm.

Additional

guidance for setting

and maintaining at least

one alarm channel

in service

for the containment

evacuation

alarm and the high flux at

shutdown

alarm

is specified in 3/4-0SP-059.6.

This item is closed.

(Closed)

IFI 250,251/85-24-05.

Determine

adequacy

of

procedures

for

making temporary

changes

per

TS 6.8'.

AP 0109.7, Responsibilities

of the

Procedure

Upgrade

Program

Group,

and

AP 0109.3,

On the

Spot

Changes

to

Procedures,

were

reviewed

and determined

to adequately

comply with the

review and approval

requirements

of TS 6.8.3.

This item is closed.

(Closed)

URI 250,251/85-26-06.

Evaluate advisability of rescaling interim

power

range

currents.

OP

0204.5,

Nuclear

Design

Check

Tests

During

Startup

Sequence

After

Refueling,

specifies

performing

a

Nuclear

Instrumentation

System

(NIS) 'detector

mini-calibration

per

OP-12304.9

prior to exceeding

50K power.

The data

obtained

are not required to be

used

for resetting

NIS

voltages

and

currents.

However, if tilt

calculations (per

ONOP 12308.2)

exceed

TS limits and flux map data is not

used

to prove that

an actual tilt condition

does

not exist,

then

the

requirements

of

TS

3'.6

(h)

and 3.2.6 (i) will be

implemented,

as

applicable.

This item is closed.

(Closed)

IFI 250,251/85-30-02,

Determine if new equipment

is promptly

added to calibration program.

Administrative Procedure

(AP) 0190.15

step

3.4. 14, requires

a meeting

coordinated

by the Engineering

Department

to

review PC/Ms for operability and maintainability.

Step 5.8.4 requires the

PC/M

coordinator

to

implement

the

required

maintenance/calibration

schedule

in

the

General

Equipment

Management

program.

This

item is

closed.

(Closed)

IFI 250,251/84-09-04,

Failure to implement

an adequate

post trip

reviews.

AP-0103. 16, Duties

and Responsibilities

of the Shift Technical

Advisor, Dated March 10,

1987, Appendix B, contains

adequate

guidance for

performing

a post trip review.

Additionally, the trip review requires the

Plant Manager - Nuclear to give permission for unit restart.

This item is

closed.

(Closed)

Deviation

250/84-04-07,

Failure to fully implement

TMI item

I.C.6,

Independent

Verification.

Procedure

O-ADM-031,

Independent

Verification dated

June

25,

1987,

provides

plant policy

and detailed

direction

on the implementation of independent 'verification requirements.

This item is closed.

(Closed)

IFI

250/84-26-01,

Reactor

,coolant

system

(RCS)

leak

rate

calculation

did

not

address

RCS

temperature

and

pressurizer

level.

Procedure

3/4-OSP-041. 1,

RCS

Leak Rate Calculation,

dated

May 29,

1987,

has

been revised to include

RCS temperature

and pressurizer

level.

This

item is closed.

(Closed) IFI 250, 251/80-06-03,

Residual

heat

removal

system

(RHR) suction

isolation

valves

MOV 750

and

751

are

not

environmentally

qualified.

Additionally, the

licensee

has

not prioritized the

RHR recirculation

switchover sequence.

The

MOVs 750 and 751 are

considered

environmentally

qualified (EQ) and are listed on the licensee's

EQ list.

The licensee

has

a

new

emergency

operating

procedure

EOP-ES-1.3,

Transfer

to

Cold

Leg

Recirculation,

which prioritizes the

RHR switchover

sequence.

This item

is closed.

(Closed)

IFI 250,

251/80-06-04,

Provide

adequate

training for management

personnel

in the area

of accident

analysis.

The licensee

has

conducted

training in the area of Mitigating Core

Damage,

which is TMI item II.B.4,

this item was closed

in inspection

report

250,

251/81-33.

This item is

closed.

(Closed)

IFI 250/84-18-03,

Review the adequacy

of the piping and supports

associated

with the containment

instrument air lines for both units.

The

licensee

completed

an

evaluation

dated

August 16,

1985

and

noted that

seismic

boundary

anchors

were

needed

to meet the current

standards

to

isolate the safety related portions of the piping from non-safety related

portions.

The licensee

has

completed

the installation of these

anchors,

Unit

3

on

May 29,

1987

and Unit 4

on February

15,

1986.

This

item is

closed.

10=

(Closed)

IE Circular 81-13,

Torque

Switch Electrical

By Pass Circuit for

Safeguard

Service

Valve Motors.

This circular required

the licensee

to

verify that all valves

important to safety

which have

the torque switch

bypass circuits installed

do in fact have these circuits and to establish

controls

to

assure

that

torque

switch

bypass

circuits

are

not

inadvertently

removed.

This circular is administratively

closed

and the

completion

of the circular's

requirements

will be tracked

under

the

response

to IE Bulletin 85-03,

Motor-Operated

Valve

Common

Mode Failure

During Plant Transients

Due to Improper Switch Settings.

This item is

closed.

(Closed) IFI 250/84-39-05

and 251/84-40-04,

The adequacy

of IEC plant work

order documentation

and procedural

guidance.

The licensee

has implemented

new procedures

that provide guidance.

These

procedures

are

O-GMI-102. 1,

Troubleshooting

and

Repair Guidelines,

and

O-ADM-701, Plant

Work Order

Preparation.

This item is closed.

(Closed) IFI 250, 251/84-09-02,

Failure to take prompt corrective actions.

The action taken

by the licensee for violation 250, 251/84-04-02,

and the

implementation

of the

PEP

program

should

provide

adequate

guidance

on

taking prompt corrective actions.

This item is closed.

(Closed)

IFI 250,

251/84-04-03,

Corrective action for Leeds

and Northrup

Speedomax

chart recorders.

The licensee

has

implemented

controlled plant

work order (84-30,

84-31) to replace

the records capillary

system with

disposable

marker s.

This item is closed.

(Closed)

IFI 250,

251/84-09-06,

Review of AFW and Air/Nitrogen System.

The licensee

completed

a review of the

AFW system,

letter dated

May 15,

1984,

Subject

Auxiliary

System

Improvement

Project.

Additionally, the

AFW system

was included in the licensees

PEP p~ogram

and

the phase II select

system review.

This item is closed.

(Closed)

URI

250,

251/86-25-06,

Determine

the

basis

for

allowing

maintenance

activities which can affect the performance of safety-related

equipment to begin without requiring that the maintenance

be preplanned

or

performed

in accordance

with written procedure.

Procedure

AP-0190. 19,

Control of Maintenance

on Safety

Systems,

requires

a

PWO

be initiated for all

maintenance

work and that

work

performed

for

emergency

situations

be

thoroughly

documented

by the

journeyman.

This item is closed.

(Closed) IFI 250, 251/85-02-06,

Research,

document

and then set the torque

switch and limit switch setting for all motor operated

valves.

This item

will

be

administratively

closed

and

action

tracked

under

IEB 85-03,

Motor-Operated

Valve

Common

Mode Failure

During Plant Transients

Due to

Improper Switch Setting.

This item is closed.

(Closed)

URI 251/86-06-03,

Evaluate the probable

cause of the misalignment

of the

4B

containment

spray

pump.

On July 16,

1986,

the

licensee

completed

an

evaluation

of the misaligned

containment

spray

pump

and

determined

the following.

The

4B containment

spray

pump shaft failed as

a

result of not verifying pump rotation

and not performing

a realignment

prior to conducting

surveillance

testing.

The

misalignment

may

have

resulted

due to improper installation

on

a pipe elbow that

had

minimum

wall thickness.

Procedure

MP 4207.2,

Containment

Spray

Pump

Disassembly,

Repair

and Assembly,

was revised to include

hand rotation of the

pump and

alignment

of

pump to motor prior to running

the

pump.

This

item is

closed.

Onsite

Followup

and In-Office Review of Written

Reports

Of Nonroutine

Events

(92700/92712)

0

The

Licensee

Event

Reports

(LERs)

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 fully completed

was clearly

assigned,

safety

questions

had

been

evaluated

and resolved,

and violations of regulations or TS conditions

had

been identified,

(Closed)

LERs

250/84-19

and

250/84-20,

TS-RCS

Leakage.

These

two

LERs

were generated

as

a result of excessive

RCS leakage that caused

two Unit 3

shutdowns.

The root cause of the events

was failed gland flanges

on seven

3/4

inch

Rockwell-Edwards

stop

valves.

The flanges

were replaced

with

carbon

steel

strong

backs

via

PC/M 84-129.

Technical

correspondence

PTN-Tech-87-182,

to the maintenance

department

specifies

maximum vendor

torque

values for the flange bolts to prevent

overtorquing

which would

lead

to intergranular

stress

corrosion cracking'ERs

250/84-19

and

250/84-20

are closed.

(.Closed)

LER 250/85-25,

Appendix

R Safe

Shutdown

Review.

This

LER was

generated

by the

licensee

to

make

advanced

notification to the

NRC of

preliminary results

of the Unit

3 Appendix

R safe

shutdown

review.

The

Region II Appendix

R inspection findings are documented

in IE Report 250,

251/86-09.

LER 250/85-25 is closed.

(Closed)

LER 250/85-33,

LOCA Analysis

Discrepancy.

This

LER

was

a

voluntary report

made

by the licensee to advise the

NRC of a discrepancy

between

a Westinghouse

(W)

LOCA analysis

and the Turkey Point

FSAR.

The

W

analysis

assumes

the failure of one of the four High Head Safety Injection

(HHSI) pumps,

the

FSAR assu'mes

the failure of two HHSI pumps.

A W safety

evaluation

reported that the

Emergency

Core Cooling System

(ECCSQ safety

criteria

as

stated

in

10 CFR 50.46,

would not

be

impacted

by the

FSAR

scenario of two failed HHSI pumps.

LER 250/85-33 is closed.

(Closed)

LER 251/85-05,

Engineered

Safety

Feature

(ESF) Actuation-Safety

Injection.

On

February 7,

1985,

following

a Unit

4 trip,

a

spurious

safety

injection

signal

was

generated

in the

ESF

system.

No safety

12

injection flow was delivered

to the

RCS.

All equipment

actuated

and

functioned

as designed.

The root cause of the event was

a blown fuse

on

a

flow comparator

(FC-485) of the

B steam generator

(SG) coincident with an

electrical

spike in the circuitry of a flow comparator

(FC-475) of the A

SG.

The blown fuse

was

replaced

and instrument calibration

checks

were

performed

on A SG flow instrumentation.

LER 251/85-05 is closed.

(Closed)

LER 251/85-07,

TS-Containment

Spray

Pump (CSP).

On February

18,

1985, with Unit 4 at

100% power, the

4A CSP was declared

inoperable.

The

4A

CSP

480V power supply breaker closing springs were discharged

and the

closing spring charging motor was turned off.

The

pump beaker

could not

have

closed

in response

to

a start signal.

It was

presumed

that this

condition

had existed

since

the last operability surveillance

test

was

performed

on

February

6,

1985.

Power

was

immediately

restored

to the

closing

springs

and

an operability test

was

performed

on

the

pump.

Operations

personnel

were

briefed

on

the

significance

of the

event.

LER 251/85-07 is closed.

(Closed)

LER 251/85-09,

TS-EDG.

On April 25,

1985, with Unit 4 at

100%

power,

the

A

EDG

was

taken

out of service

for periodic

maintenance

coincident with the

3B 4160V bus being out of service.

This rendered

the

3A,

4A and

3B high

head

safety injection

pumps

inoperable,

which is in

noncompliance

with TS.

Upon recognition of this

event,

the

A

EDG

was

tested

and returned to service.

Operations

personnel

were briefed

on the

importance

of ensuring

the operability of opposite

train

ESF

equipment

prior to electively

removing

an

EDG

from service.

LER 251/85-09

is

closed.

(Closed)

LER 251/85-18,

TS Heat Tracing.

On June

23,

1985, with Unit 4 at

27% power,

two channels

(8A 5 8B) of critical heat tracing

on the boric

acid

pump suction lines were declared

out of service.

TS allow only one

channel

to be inoperable.

A plant shutdown

was implemented

as required

by

TS 3.0. 1.

The root

cause

of the failed

channels

was

a

short circuit

created

when

excess

heat tracing wiring from circuit 9 contacted circuit

8.

Both circuit 8 channels

were

repaired

and

returned

to

service.

Circuit

9

was

shortened

to

prevent it from contacting

circuit 8.

LER 251/85-18 is closed.

(Closed)

LERs

251/84-09

and 251/85-20,

TS-Containment Integrity.

These

LERs were

generated

as

a result of events

in which Operations

personnel

didn't fully appreciate

the

TS

requirements

for containment

integrity.

Later

similar events

led to

issuance

of violation 251/86-41-01.

The

corrective

actions

to this violation,

as

stated

in

the

licensee's

response,

address

the corrective actions of the

LERs.

LERs 251/84-09

and

251/85-20 are closed.

(Closed)

LER 250/85-39,

On

November 29,

1985,

the Engineering

Department

notified Turkey Point that portions of the accumulato~ fill line were not

seismically installed.

PC/Ms 86-80 and 86-004,

Unit 3 and

4 Accumulator

Makeup

Header - Seismic. Upgrade,

have

been

completed

by the

Nuclear

Startup

Department

and turned over to Operations.

This item is closed.

13

(Closed)

LER 251/86-08,

On

March 29,

1986,

the

4A Intake Cooling Water

( ICW)

Pump

Was

Inadvertently

Started.

The

actuation

was

due

to

a

construction

worker physically disturbing

the relay.

The construction

worker received

instructions

to

use

caution

when working in the vicinity

of safeguards

equipment.

This item is closed.

(Closed)

LER 250/85-34,

On October 23,

1985, the

3A Residual

Heat

Removal

(RHR)

Pump

was declared

out of service

due to failure to meet

the

seal

leakage

acceptance

criteria during

an operability test.

The

licensee

repaired

the pump's mechanical

seal

and conducted

a post maintenance test.

This item is closed.

(Closed)

LER 250/85-40,

On

November

11,

1985,

Unit 3 subcritical

reactor

trip occurred

as

a result of manually re-inserting nuclear instrumentation

system

channel

N-32 instrument

power fuses while attempting to energize

the channel.

The licensee

repaired the voltage

power supply

and replaced

a

capacitor

in

the

channel

pre-amplifier.

Additionall'y,

all

post-maintenance

testing

was

satisfactorily

completed.

This

item is

closed,

(Closed)

LER 250/85-41,

On

December

4,

1985,

AFW initiated

due to

an

improper

alignment

of the

condensate

system.

The

licensee

properly

aligned the condensate

system,

the operators

involved were

counselled

on

the

need

for clear

and

concise

inter shift turnovers

and

procedure

3/4-0P-073,

Condensate

System,

was revised to require

a plant clearance

for the

Steam

Generator

pump motor breaker

when the condensate

system

was

aligned

in the recirculation

cleanup

mode.

This

item is

closed'Closed)

LER 250/86-25,

On June

12,

1986,

the

3A and

3C charging

pumps

were out of service,

exceeding

TS requirements.

The

3C pump was repaired

and

retested

satisfactory.

The

3A

pump

cracked

weld was repaired

and

satisfactorily

placed

back

into

service.

Additionally, the

licensee

developed

a

PC/M to replace

the

pump packing with longer life packing.

This item is closed.

(Closed)

LER 251/86-09,

On April 10,

1986, it was determined that

a flow

reversal

condition existed

concerning

the

component

cooling water

(CCW)

supply to high head safety .injection (HHSI) pumps

seal

and thrust bearing

coolers.

The

licensee

has

completed

PC/M

83-008,

to correct

piping

misrouting.

This item is closed.

(Closed)

LER 251/86-17,

On August 9,

1986, the Unit 4 Auxiliary Feedwater

System

was

actuated

during

a

system test

due to personnel

error.

The

operator

received

counselling

concerning

his

actions.

This

item is

closed.

(Closed)

LER 251/86-19,

On

September

6,

1986, while Unit 4 was at

38K

power,

a reactor trip occurred

due

to

a

4C

steam

generator

isolation circuity failure.

The failed circuity,

a light socket,

was

replaced

and the other sockets for Unit 4 were inspected.

The turbine

14

trip solenoid

was replaced

and procedure

3/4-0SP-089,

Main Turbine Valves

Operability Test,

was revised to test the turbine trip solenoids.

This

item is closed.

(Closed)

LER 251/86-21,

Unit 4, Shutdown

on September

16,

1986,

due to Rod

Position Indication (RPI)

System

Malfunction.

The failed line voltage

regulator

was replaced.

The licensee

developed

a preventative

maintenance

procedure,

0-PME-028. 1,

RPI Inverter Maintenance.

This item is closed.

The following LERs were reviewed

and closed

based

on an in-office review.

The

inspectors

verified that reporting

requirements

had

been

met, root

cause

analysis

was performed, corrective actions

appeared

appropriate,

and

generic applicability

had

been

considered.

In addition,

each

LER was

reviewed for 'and

determined

not

to

require

further

onsite

inspector

followup.

LER 250/85-22,

Reactor Protection Actuation Reactor Trip

LER 250/85-31,

Engineered

Safety Features

Actuation - Turbine Runback

Monthly and Annual Surveillance

Observation

(61726/61700)

The

inspectors

observed

TS required surveillance

testing

and verified:

that the test

procedure

conformed to the requirements

of the

TS, that

testing

was

performed

in accordance

with adequate

procedures,

that test

instrumentation

was calibrated,

that limiting conditions

for operation

( LCO) were met, that test results

met acceptance

criteria requirements

and

were reviewed

by personnel

other than the individual directing the test,

that

deficiencies

were identified,

as

appropriate,

and

were properly

reviewed

and resolved

by management

personnel

and that system

restoration

was

adequate.

For completed tests,

the inspectors

verified that testing

frequencies

were met and tests

were performed

by qualified individuals.

The

inspectors

witnessed/reviewed

portions

of

the

following test

activities:

Unit 3 Engineered

Safeguards

Integrated Test,

3-OSP-203

Auxiliary Feedwater Train

1 Operability Verification, 4-OSP-075.

1

Nuclear Plant Operator

Logsheets,

4-0SP-201.3

System

Flowpath Verification, 4-0SP-075.5

Safety Injection

Pumps Inservice Test,

O-OSP-062.2

No violations or deviations

were identified within the areas

inspected.

Maintenance

Observations

(62703/62700)

Station

maintenance

activities of safety related

systems

and

components

were

observed

and

reviewed

to ascertain

that

they

were

conducted

in

accordance

with approved procedures,

regulatory guides,

industry codes

and

standards

and in conformance with TS.

15

The following items

were considered

during this review,

as appropriate:

that

LCOs were met while components

or systems

were removed

from service;

that approvals

were obtained prior to initiating work; that activities

were

accomplished

using

approved

procedures

and

were

inspected

as

applicable;

that

proce'dures

used

were

adequate

to control the activity;

that

troubleshooting

activities

were

controlled

and

repair

records

accurately

reflected

the maintenance

performed; that functional testing

and/or

calibrations

were

performed

prior to

returning

components

or

systems to service; that

QC records

were maintained; that activities were

accomplished

by qualified personnel;

.that parts

and materials

used

were

properly certified; that radiol,ogical controls were properly

implemented;

that

QC hold points were established

and

observed

where required;

that

fire prevention controls

were

implemented;

that outside contractor

force

activities were controlled in accordance

with the'pproved

QA program;

and

that housekeeping

was actively pursued.

The following maintenance activities were observed

and/or reviewed:

Intake Cooling Water

Pump Motor - Overhaul

and Maintenance,

MP 3407.6

Pump

C Steam

Leak Repair,

PWO 69-2507

Emergency Diesel Generator

Pump

C Trip

and

Throttle

Valve

Repair,

PWO

69"5668

A Skid 'Tank Leak Repair,

PWO 300812

Emergency Diesel

Generator

A Governor Troubleshooting

Steam Trap ST-53 Drip Leg Drain Valve Repair,

PWO 019

No violations or deviations

were identified within the areas

inspected.

However, violation 251/87-33-02 discussed

in paragraph ll appears

to have

resulted

due

to

maintenance

personnel

failing to

inform appropriate

managers that

a suspected

valve packing leak was observed to originate at

a degraded

weld in the train

1 auxiliary feedwater

system.

Operational

Safety Verification (71707)

The inspectors

observed control

room operations,

reviewed applicable logs,

conducted

discussions

with control

room

operators,

observed

shift

turnovers

and confirmed operability of instrumentation.

The

inspectors

verified the operability

of selected

emergency

systems,

verified that

maintenance

work orders

had been

submitted

as required

and that followup

and prioritization of work was

accomplished.

The

inspectors

reviewed

tagout records, verified compliance with TS

LCOs

and verified the return

to

service

of affected

components.

Additionally, by observation

and

direct interviews, verification was

made that the physical

security

plan

was

being

implemented.

Plant

housekeeping/cleanliness

conditions

and

implementation of radiological controls were also observed.

Tours of the

intake structure

and diesel, auxiliary, control

and turbine buildings were

conducted

to observe

plant equipment

conditions including potential fire

hazards, fluid leaks

and excessive

vibrations.

16

The

inspectors

walked

down accessible

portions of the following safety

A and

B Emergency Diesel Generators

Control

Room Vertical Panels

and Safeguards

Racks

Intake Cooling Water Structure

4160 Volt Buses

and 480 Volt Load and Motor Control Centers

Fire Protection

Deluge Valves

a.

Technical

Specification

Requirements

Not

Implemented

For Reactor

Protection Setpoint Reductions

On

July 12,

1987,

dur ing

routine

backshi ft

inspection

at

approximately 7:00 p.m., while the Unit 4 reactor

was at

76 percent

power,

the inspector

determined

that that the licensee

was

not in

compliance with TS 3.2.6.(i).

This specification

requires,

in part,

that

when the Quadrant

Power Tilt Ratio

(QPTR)

exceeds

two percent

for

24

hours

and

the

reactor

hot

channel

factors

have

not

been

determined

to be acceptable,

then reactor protection

setpoints for

Over-Power

Differential

Temperature

(OPDT)

and

Over-Temperature

Differential Temperature

(OTDT) shall

be reduced.

QPRT is a measure

of radial

power differences existing in the upper and lower quadrants

of the

reactor

core

as

measured

by the

four excore

power

range

nuclear instruments

(PRNIs).

The licensee

took immediate

action to verify that the hot channel

factors

were acceptable

by performing

and evaluating

a Unit 4 flux

map

using

moveable

incore

detectors.

Subsequent

evaluation,

completed

on July 13,

1987 at approximately

1:50 a.m., verified that

the hot channel

factors were acceptable.

This alleviated the need to

reduce

the

OPDT and

OTDT setpoints.

The incore detectors

indicated

that the quadrant

power ratios did not exceed

two percent.

The

QPTR annunciators

are,

by design,

inoperable

below

50 percent

reactor

power.

Standard

TS

and the licensee's

interim

TS address

QPTR in terms of compensatory

actions only when reactor

power exceeds

50 percent.

During initial reactor

startup

following refueling and

PRNI detector

replacements,

QPTR calculations

are

likely to

be

erroneous

until

the

PRNI

can

be

calibrated.

Consequently,

by

administrative

procedure,

the licensee

performs flux maps prior to

exceeding

50 percent

power to allow interim calibration of each

PRNI.

Calibration of the

PRNI at

50 percent

power

causes

the

QPTR

(as

calculated

using the excore

PRNIs) to closely agree with the quadrant

power ratios measured

during the flux mapping.

The licensee

performed

two flux maps at 50 percent

power

on July 8,

1987,

because

the detector for PRNI N-41 had been replaced during

a

recently concluded

outage.

One

map indicated

a

QPTR of 2 '

percent

for the

lower core with peaking factors

which would have

precluded

reactor operation at

100 percent

power.

This flux map was

suspected

17

of containing

erroneous

data

due to maintenance

problems

which had

been

experienced

with the

moveable

detector

drive

mechanisms.

A

second

flux

map

indicated

no

QPTR

greater

than

2

percent.

Additionally, the results of this second

map compared favorably with

similar maps performed prior to the outage.

Calibration data derived

from the flux maps

was

not installed

in the

PRNIs

because

one

instrument,

N-44, was out of service.

Taking

a second

instrument out

of service for calibration would result,

by design,

in

a reactor

trip.

Since

calculated

gPTR

was

less

than

two

percent,

power

escalation

was'uthorized

by the

Reactor

Engineering

Supervisor.

Power

was

slowly increased

between

July 10-11.

QPTR calculations

performed in the morning

and

evening

on July 10 were less

than

two

percent.

However,

a

gPTR calculation

which was

begun

on July 10 at

ll:50 p.m.

and completed

on July ll at 12:Ol a.m. indicated

an

upper

detector tilt of 2.84 percent.

The increase

in gPTR with increasing

power was not expected

and was

not clearly understood.

Consequently,

when the

Reactor

Engineering

Supervisor

was

contacted

at

home

and

informed of the

change,

he

recommended

that reactor

power be maintained

below full power by two

percent

for each

1 percent

of

gPTR.

This action

implemented

the

requirements

of

TS 3.2.6.(h).

gPTR

was calculated

numerous

times

between

the early morning

on Saturday July ll and the early evening

on July 12,

a

period

of approximately

43

hours.

All results

indicated that the

gPTR was greater

than two percent.

Additionally,

the magnitude of the tilt appeared

to change.

By July ll at 7:50 a.m.

it had increased

to 5.2 percent.

It then decreased

to 2.0 percent

on

July

12 at ll:30 a.m.

and began to increase

reaching

2.4 percent at

7:05 p.m.

on July 12.

Discussion with members of the Operations staff on July 12 revealed

a

lack of concern for the indicated condition.

Several

reasons

were

presented

as to why the condition was not

a concerns

These

included:

(1)

a belief that

the flux

maps

performed

on July 8

adequately

verified that the indicated

gPTR was not real; (2)

a perception that

the

gPTR was relatively stable;

(3)

a belief that the high ratio was

caused

by the inability to install corrected calibration currents in

the

PRNIs

because

instrument

N-44

was

out of service;

and (4)

a

belief that

the

safety

significance

of the

indicated

problem

was

diminished since reactor

power

had

been

maintained

below the limits

specified

in

TS 3.2.6.(h).

Additionally, the Operations

staff did

not appear

to

be

aware of the requirements

to reduce

the

OPDT and

OTDT setpoints

as specified in TS 3.2.6.(i).

Discussions

with

an

NRC core physics

inspector

confirmed that the

above reasonings,

evaluated singularly and in unison, did not provide

an engineering

basis

upon which to verify that the indicated

gPTR

condition

was not real.

Such

a verification would generally

have

been

accomplished

through the performance of additional flux maps at

the power level

and control rod position that existed

when the

gPTR

exceeded

two percent.

The licensee

declined to perform additional

18

flux maps

even though the reactor

had been

increased

from 50 percent

to

90 percent

power

and

had suffered

a small turbine

runback since

the last flux maps

were taken.

This option is authorized

by TS

as

long

as all appropriate

compensatory

power reductions

and reactor

protective setpoint reductions

are

implemented.

This failure to

implement

the

requirements

of

TS 3.6.2.(i) is

a

violation.

This violation applies to Unit 4 only (251/87-33-03).

Licensee

Determined

That

The Auxiliary Feedwater

(AFW) System

Was

Inoperable

On July 15,

1987, with the Unit 4 reactor at

100 percent

power, the

licensee

determined that the Unit 4 AFW system

was inoperable

because

both trains of safety-related

nitrogen for the automatic flow control

valves

were

isolated.

The

discrepancy

was

discovered

during

a

routine inspection

by

a Turbine

Operator

tasked

with logging the

status

of the

nitrogen

system

on

a periodic (4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) basis.

The

discrepancy

was

immediately

reported

to the

Operations

Department

staff

and the erroneously

closed

valves

were

opened,

returning the

system

to

service.

"Procedure

4-0P-065.2,

and

Main

Steam

Isolation

Valve

(MSIV) Nitrogen

Gas

Supply

System,

revision dated

June

18,

1987,

was expeditiously performed to verify that all valves

were in the correct position.

The discrepancy

received the highest levels of management

attention,

including the establishment

of a

human

performance

review

team

to'nvestigate

and identify the root cause

of the valve misalignments.

It was determined that valves which were required to be

open

had been

closed earlier

on July 15,

1987 by a Turbine Operator.

The operator

had

previously

been

assigned

for

an

extended

period of time to

Unit 3, which was in

an

extended

refueling outage.

The

AFW system

was not required

by

TS to

be

operable

on Unit 3

because

of the

shutdown

status

of the

reactor.

Apparently the

operator

was

not

aware that

a June

1987 revision to procedure

4-0P-065.2

required

3

nitrogen bottles

per train

be in service,

as

opposed

to

one bottle

which

had

previously

been

acceptable.

During

his

1:00 a.m.

inspection of the nitrogen bottles

he noticed that three bottles were

in service

instead

of one.

Believing this to be

a discrepancy,

he

realigned

the bottles with out the

use of the

approved

procedure.

This caused all bottles to be isolated, either by the valve on top of

the bottle or the in-line isolation valve further down stream being

closed.

This valve realignment

was

performed for both

AFW nitrogen

trains,

rendering both inoperable.

The perceived

discrepancy

was not reported to the Operations staff by

the Turbine Operator.

No quality record was created to document

the

realignment.

Logs taken

on the nitrogen

system status at 5:00 a.m.,

9:00 a.m.,

and 1:00 p.m.

documented

(erroneously)

that

1 bottle

was

in service

per train.

The 5:00 p.m.

log reading

indicated that

no

bottles

were

in service

but this

fact

was

not

brought

to

the

19

attention

of supervisory personnel.

The Turbine Operator taking the

9:00 p.m.

log readings identified the discrepancy.

TS 3. 18 requires that the

AFW be operable

when the reactor is heated

above

350 degrees.

One train is allowed to be out of service for 72

hours.

Two trains

are

not allowed to

be

simultaneously

out of

service.

The

Turkey

Point

system

design

basis

document,

developed

under the

Performance

Enhancement

Program,

specifies that

the

AFW system

shall

be capable of automatic operation

upon loss of

instrument air for

a period of

two

hours

without

any

required

operator

action

outside

the control

room.

The

AFW flow control

valves

normally use

the non-safety

related,

non-seismic

instrument

air

system

for automatic

valve positioning.

The

instrument air

system

can not,

because

of its unqualified

nature,

be

assumed

to

exist for post accident

AFW operation.

The nitrogen

system,

although

it is often

referred

to

as

a

backup

system,

is required

to

be

operable

to

support

post

accident

AFW system operation.

For this

reason,

the

nitrogen

system,

like

the

system,

is

both

safety-related

and seismically installed.

During the 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months that the nitrogen trains were out of service the

instrument air system

remained in service

and operated

normally. If

instrument

air

pressure

had

become

excessively

low,

existing

administrative

procedures

require that the reactor

be shutdown.

The failure to implement the requirements

of TS 3. 18,for AFW system

operation is

a violation which applies to Unit 4 only (251/87-33-01).

Inadequate

Procedure

For Deluge Fire Systems,

Repeat Violation

On July 14,

1987,

the inspectors

performed

a walkdown of the fire

suppression

systems

for the Unit 3

and

4

component

cooling water

(CCW) pump rooms.

One Unit 3 deluge valve,10-837,

was not correctly

aligned in that

a pressure

switch was isolated.

This discrepancy

did

not prevent

the deluge

valve from operating

but it did prevent the

activation

of the control

room

and

local

area

actuation

alarms.

Since the pressure

switch was isolated,

the 'activation of the deluge

station

would not

be received at the control

room

and local

alarm

panels.

Indication of fire protection

system actuation

would be only

by

secondary

means

such

as

low fire main

pressure

or fire

pump

initiation.

The ability

to

remotely

verify flow through

the

appropriate

deluge station would be lost.

This discrepancy

is

a

repeat

of

a similar problem

described

in

Inspection

Report

250,

251/86-33 dated

September

3,

1986.

Violation

250,251/86-33-03

was

issued

because

adequate

procedures

for the

control

of the

deluge

valves

did

not exist,

contrary

to

the

requirements

of TS 6.8.1.

The licensee

responded

to the Notice of

Violation

on

October 3,

1986

in

PTN-TECH-86-743.

The

proposed

corrective action was to revise

procedure

0-OP-016. 1, entitled Fire

Protection

Water System,

to incorporate

deluge

system valve lineups.

20

Procedure

0-OP-016. 1 was revised

on

December 9,

1986 to include the

auxiliary

support

valves

necessary

to properly align the

deluge

system,

However, for each deluge

system,

the isolation valve for the

alarm

pressure

switch

was

omitted

from

the

lineup

sheets.

Consequently,

the procedure

remained

inadequate

because

the pressure

switches

for the

alarm stations

remain

isolated

when

the

deluge

systems

are returned to service.

TS 6.8. 1 requires that written procedures

and administrative policies

be

established

that

meet

or

exceed

the

requirements

and

recommendations

of Appendix

A of USNRC Regulatory

Guide 1.33.

Regulatory

Guide 1.33,

Appendix A, states

that procedures

should

be

established

for the operation of plant fire protection

equipment.

The fai lure to have

an

adequate

procedure for the alignment of fire

protection deluge

systems is

a repeat of violation 250,

251/86-33-03

(250, 251/87-33-04).

11.

Engineered

Safety Features

Walkdown (71710)

To verify system operability the inspectors

performed

a complete walkdown

of all accessible

equipment

of the Unit

4 Auxiliary Feedwater

(AFW)

system.

One train of steam to the

AFW pumps

was found to be degraded

as

described

below.

This matter was immediately brought to the attention of

the

licensee

and

NRC

Region II management.

The licensee

declared

the

train inoperable

and

implemented

the

shutdown

requirements

of TS 3.0.1.

Unit 3 remained in cold shutdown

(mode 5) during the inspection period and

consequently its Engineered

Safety

Features

were not required to

be in

service.

The following criteria

were

used,

as appropriate,

during the

walkdown:

a.

System

lineup

procedures

matched

plant

drawings

and

the as-built

configuration.

b.

Equipment conditions

were satisfactory

and

items that might degrade

performance

were identified and evaluated

(e.g.

hangers

and supports

were operable,

housekeeping

was adequate).

c.

Instrumentation

was

properly valved

in

and functioning

and that

calibration dates

were not exceeded.

d.

Valves were in proper position, breaker alignment was correct,

power

was available,

and valves were locked/lockwired as required.

Local

and

remote

position

indication

was

compared

and

remote

instrumentation

was functional.

f.

Breakers

and instrumentation

cabinets

were

inspected

to verify that

they were free of damage

and interference.

21

A walkdown of the Unit 4 portion of the

AFW system

was performed

between

July

17 and July 20,

1987.-

On July 17, plant parameters

were

steady

and

no

demand for AFW system

operation

existed.

The reactor

was critical at

less

than

1 percent

power and the turbine generator

had been

removed from

service

due to apparent

condenser

tube leakage.

On July

17

the

inspector

observed

a pinhole

leak in the

steam

supply

piping which supplies

turbine driven

AFW pumps

A and

C.

The

leak

was

located

on

a

two inch diameter

pipe that branches

from the four inch

diameter train

1 steam line.

The two inch pipe supplies

steam trap

53 and

system

low point drain

AFSS 43.

The leak was located adjacent to AFSS 43.

Additionally, the

2 inch diameter pipe was heavily corroded.

The licensee

determined that

a maintenance

concern

had been identified on

July 11,

1987,

when plant personnel

reported

seeing

small

amounts of water

drop

from the vicinity of drain

valve

AFSS

43.

This resulted

in the

issuance

of a deficiency tag

and Plant Work Order

(PWO) which stated that

either

AFSS

43

had

a valve

stem

packing

leak or the

steam

pipe

was

leaking,

A definitive evaluation

of the source of the leak could not be

made

because

the area of concern

was obscured

by insulating lagging.

The

PWO

was

erroneously

classified

as

Non-Nuclear Safety

and

therefore

did not receive

high priority.

Troubleshooting

did not begin

until the morning of July 14,

1987,

when the

lagging

was

removed

and

a

Journeyman

confirmed the existence

of the corroded

pipe

and the pinhole

leak.

Between

July 14-17,

a

weld repair

plan

was

developed

by the

Mechanical

Maintenance

Department.

No evaluation of the

extent of the

degradation

was performed

and

no operability assessment

of the Train

1

steam

supply

was

performed.

The

steam

header

remained

in service

and

aligned

for

automatic

operation.

The

PWO

remained

classified

as

non-safety related.

In the early

afternoon

on July 17,

the

inspector

requested

that the

licensee

evaluate

the extent of the pipe corrosion

and its effect on

10 CFR 50, Appendix B, Criterion XVI, as

system operability.

The licensee

determined that the piping was actually

safety

a

Non-Conformance

Report

(NCR)

was

issued

to the

Engineering

Department for evaluation.

At approximately 4:30 p.m. it was

determined that,

based

on the extensive visible corrosion

and the observed

pipe leak, the integrity of the

steam line was

suspect

and could not be

assured

without further testing.

A specific concern existed that the weld

joint at which the pinhole leak originated

was of unknown quality and that

pipe wall thinning in excess

of allowed tolerances

could

have resulted

from the observed corrosion.

The licensee

determined

that the train

1

steam line was

inoperable

and

that sufficient evidence

of the inoperable

condition

had

been

obtained

during the July

14 pipe inspection

and

was

not effectively evaluated.

Consequently,

the licensee

determined

that

AFW train

1

had

been

out of

service in excess

of the

72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Limiting Condition for Operation

(LCO)

allowed by TS 3. 18 for single train operation.

22

TS 3.0. 1 requires,

in part, that when

an

LCO is not met then the reactor

shall

be

placed

in hot

standby

(mode

3) within six hours.

Licensee

interviews with the Journeyman

revealed that

he started

work on July 14 at

7:00 a.m. but his recollection

was that

he did not remove the lagging and

expose

the

source

and nature of the leak until approximately ll:00 a.m..

Consequently,

the licensee

determined

that the

72

hour

LCO expired

at

11:00 a.m.

on July

17

and

mode

3 was required at 5:00 p.m.

At 5:12 p.m.

on July 17,

1987 the Unit 4 reactor

was placed

in hot standby

and plans

were

implemented

to bring the unit to cold shutdown

(mode 5) as required

by additional portions of TS 3.0. 1.

This

event

is of concern

because

the

licensee

identified

a condition

adverse

to quality

and initially failed to

recognize it

as

such.

Maintenance

troubleshooting

and repairs

were

erroneously

identified

as

non-safety

related.

This

resulted

in

a

three

day delay

before

the

potential

through wall pipe leak postulated

in the

PRO

was investigated

(July 11-14).

An appreciation for the potential of train failure through

weld degradation

and/or pipe wall thinning was not shown

by the Mechanical

Maintenance

Oepartment

subsequent

to the July

14 inspection.

implemented

by Florida Power and

Light Topical Quality Assurance

Report

FPLTQAR 1-76A, Revision

10,

and

TQR

16.0,

Revision

5, entitled Corrective

Action, requires,

in part, that

measures

be established

to assure

that conditions adverse

to quality,

such

as failures,

malfunctions,

deficiencies,

deviations,

defective material

and equipment,

and nonconformances

are promptly identified and corrected.

FPL

Quality

Assurance

Manual,

Quality

Procedure

16. 1,

Revision

8,

delineates

requirements

for assuring

that conditions

adverse

to quality

are promptly corrected.

The failure to promptly identify and correct

a condition

adverse

to

quality

is

a

violation.

This

violation

applies

only

to

Unit

4

(251/87-33-02).

Plant Events (93702)

The following plant events

were reviewed to determine facility status

and

the

need for further

followup action.

Plant

parameters

were evaluated

during transient

response.

The significance of the event

was evaluated

along with the

performance

of the

appropriate

safety

systems

and

the

actions

taken

by the

licensee.

The

inspectors

verified that required

notifications were

made to the

NRC.

Evaluations

were performed relative

to the

need for additional

NRC response

to the event.

Additionally, the

following issues

were

examined,

as

appropriate:

details

regarding

the

cause

of the event;

event chronology; safety

system performance;

licensee

compliance

with approved

procedures;

radiological

consequences,

if any;

and proposed corrective actions.

The licensee

plans to issue

LERs on each

event within 30 days following the date of occurrence.

23

On July 1,

1987, while Unit 3 was in cold shutdown,

Unit 3 underwent

a

safety injection automatic initiation from

a

containment

high pressure

signal. Electrical

department

technicians

required

a

hose

connection

to

test penetration

canisters

on Unit 4.

To accomplish this they borrowed

an

IKC pressure

regulator

from Unit

3 being utilized in preparation

for

initiating a containment

high pressure

signal for safeguards

testing.

The

high pressure

nitrogen

supply

was manipulated

to verify valve closure.

This caused

a high containment

pressure

signal.

On July 5,

1987,

while Unit

3

was

in cold

shutdown, the

D-MCC [Motor

Control

Center]

was deliberately

de-energized

during Unit 3 Integrated

Safeguards

testing.

D-MCC supplies

power to the

4A emergency

containment

cooler

fan and'ssociated

CCW valves.

On loss of power, these

associated

CCW valves failed open

as designed,

causing

Unit 4

CCW flow to increase

which

lowered

header

pressure.

This

caused

the

4B

CCW

pump automatic

start

on low pressure

signals

On July 15,

1987,

due to

an incorrect valve alignment,

while Unit 4 was

operating at 1004 power, train

1 and

2 AFW backup Nitrogen was unavailable

for approximately

20

hours.

This

issue

is

discussed

in detail

in

paragraph

10.

The licensee

promptly returned the system to operation

when

the problem was discovered.

On July 17,

1987,

Unit 4 was placed in hot shutdown 'due to

a leak on

train

1

steam

supply line.

Licensee

personnel

determined that the leak,

which was very small,

existed

due to

a degraded

weld in the

AFW system

train

1 steam

header.

Since repairs

were not initiated within the allowed

72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

LCO specified in Technical Specification 3. 18, the licensee

placed

Unit 4 in mode

3 and subsequently,

mode 4,

as required

by TS 3.0. 1.

This

issue is discussed

in detail in paragraph

11.

Summary of International

Atomic Energy Agency (IAEA) Activities

In fulfillment of the

Safeguards

Agreement

between

the United States

and

the

IAEA, the

IAEA selected,

on July 19,

1985,

Turkey Point Unit 4 for

participation in its international

safeguards

inspection

program.

A major

portion of this program requires

the continuous surveillance of the fuel

inventory

through

camera

monitoring

and

seal

wire

placement.

The

surveillance

program

ensures

that

the

fuel

inventory

does

not

change

between

physical audits.

The inspectors

verified, during routine tours of the Unit 4 Spent

Fuel

Pool

(SFP)

and the accessible

portions of the containment building, that

seal

wires were in place

and intact

and that surveillance

cameras

were

operable.

Seal

wires are

placed

by

IAEA inspectors

on the containment

equipment

access

hatch,

the missile shields

and the reactor

vessel

head

seismic restraints.

Only the

seal

wires

on the equipment

hatch

can

be

observed

from outside the containment building.

The containment

building

is not normally entered during power operation.

Two surveillance

cameras

are installed

in the Unit 4

SFP.

The

SFP

area

is

always

accessible

through locked and alarmed doors.

Two

IAEA inspectors,

accompanied

by an

NRC representative,

visited the

site

on July 16,

1987.

Work performed

included

changing

the film in the

two Unit 4 spent

fuel pool monitoring cameras,

placing seal wires on the

Unit 4 equipment hatch,

and reviewing fuel inventory records.

By mutual

IAEA, NRC and licensee

agreement,

seal wires were not placed

on the Unit 4

missile

shields

because

neutron

dose

rates

near

the reactor

head

are

prohibitively high while the reactor is critical.

These

seals will be

installed during

a subsequent visit when the reactor is subcritical.

Plant Procedures

(42700)

A review

was

performe'd

of selected

plant

procedures

to verify, that

overall plant procedures

are in accordance

with regulatory requirements,.

that procedure

changes

are

made

in accordance

with TS requirements

and

that procedures

are adequate.

Numerous

procedures,

including administrative

procedures

(ADM), emergency

operating

procedures

(EOP),

off-normal

operating

procedures

(ONOP),

operating

procedures(OP),

and surveillance

procedures

(OSP) were reviewed

to verify that appropriate

reviews

and approvals

were

performed prior to

issuance.

It

was

determined

that

an

effective

procedure

review

and

approval

program exists

and is being

implemented

in

a

manner

consistent

with

TS

requirements.

These

requirements

include

review of proposed

procedures

by the Plant Nuclear Safety

Committee

(PNSC)

and approval

by

the Plant Manager-Nuclear.

Of the

40 procedures

selected

at random, all

had been

reviewed

by the

PNSC and approved

by the Plant Manager-Nuclear.

A'eview

was

performed

to verify that

procedure

changes

were

made to

reflect

TS changes

and license

revisions.

TS

amendment

numbers

118 and

112,

for Units

3

and

4 respectively,

established

TS 3.20,

Standby

Systems,

which

requires

two

standby

pumps

to

be

available with 60,000 gallons of water in the demineralized

water storage

tank.

The

licensee

implemented,

on

October

14,

1986,

procedure

0-OP-074. 1, to provide standby

feedwater operating instructions

and valve

alignment

guidance.

Additionally, surveillance

procedure

O-OSP-074.3,

Standby

Steam

Generator

Pumps

Availability Test,

has

been

developed

to implement the surveillance

requirements

specified in TS 4.21

for the

system.

Procedure

0-OSP-200. 1,

Schedule

of Plant'hecks

and

Surveillances,

revision

dated

July 17,

1987,

implements

the

monthly

requirement

to perform

O-OSP-074.3.

Procedures

also

existed

for the

verification of demineralized

water

storage

tank

level

(each shift,

OP-0204.2)

and testing

the

standby