ML17342A907

boron concentration,

in

that,

a satisfactory

sample

was not obtained

from the

4C accumulator prior

to heatup

above

200F (paragraph

8) (251/87-35-01).

Two examples

of failure to meet the requirements

of TS 6.8. 1, in that,

valve 3-40-856

was not properly controlled (locked closed)

as required

by

approved administrative

procedure

and

a compensatory

continuous firewatch,

required

by administrative

and

temporary

procedures,

was

found asleep

(paragrapl

10) (250,251/87-35-02).

Failure to meet

the requirements

of TS 3.7.2.b,'n that, the A Emergency

Diesel

Generator

(EDG) was out of service for greater

than twenty-four

hours without the

remaining

B

EDG being

tested

to

prove operability

(paragraph

10) (250,251/87-35-03).

Unresolved

Items (URI)

Unresolved

items

are matters

about which more information is required to

determine

whether

they

are

acceptable

or

may

involve violations

of

requirements

or deviations

from commitments.

One

unresolved

item

was

identified during this inspection period.

Determine'he

adequacy

of

Administrative

Site

Procedure

(ASP)-2,

revision 4,

section 6.6.9.

The

procedure

allows

process

sheets

and

process

sheet

revisions

to

be

implemented

without prior Plant

Nuclear

Safety

Committee

(PNSC)

approval,

which

may

be, contrary

to

the

requirements

of TS 6.5. 1.6.d.

(paragraph

14) (250,251/87-35-04).

I

Inspector

Fol lowup Items (IFI)

Track licensee

development of a mechanism to mark superseded

drawings kept

in document control

ready

use files such that they are not issued for use

to plant personnel

(paragraph

14) (250,251/87-35-05).

Evaluate

the

circumstances

surrounding

the

apparent

installation of an

incorrect

check valve in the Unit 3 Auxiliary Feedwater

(AFW) nitrogen

system

(check

valve '93).

- Evaluate

the circumstances'urrounding

the

removal

of spacers

between

the

Current

Pneumatic

(I/P) module for AFW

valve

FCV 3-2832 (paragraph

14) (250/87-35-06).

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.

Verification of corrective

action

was

achieved

through record reviews,

observation

and discussions

with licensee

personnel.

Licensee

correspondence

was

evaluated

to

ensure

that

the

responses

were timely and that corrective actions

were "implemented within

the time periods specified in the reply.

(Closed)

Violation 250/83-41-07

and

251/83-40-07.

Post

modification

System

Restoration

Without Complete As-Built Packages

-

3 examples.

response

letter dated

March 19,

1984 was found to

be acceptable

(per

NRC

letter dated

August

17,

1984).

The thrust of the violation dealt with

programmatic

problems

associated

with turnover

and

testing

of Plant

Change/Modifications

(PC/M).

The

inspector

reviewed

the

following

procedures

for missing aspects

indicated in the originating re'port details

section:

ASP-21

Turkey Point Plant "Startup", Revision

3.

ASP-11

Tur key Point Plant "Construction Turnover".

AP-0190.15

Plant

Changes

and Modifications (PC/M).

The procedures

contained

the aspects

referred

to in the subject report.

The specific fixes for this violation appear

to

be adequate.

The design

control

program

has

been selectively reviewed by the

NRC since

issuance

of

this violation.

Design

control activities at

the site

which

include

broader

programmatic

implementation

of the

above violation's

cor'rective

action are

encompassed

in other

NRC documents

such

as

Confirmatory

Order

EA 86"20.

Violation 250/83-41-07

and 251/83-40-07 is closed.

(Closed) Violation 250/83-41-01

and 251/83-40-01.

Failure to Compensate

Intermediate

Range

Nuclear

Instrumentation

Adequately.

response

letter dated

March

19,

1984

was

found to

be acceptable

(per

NRC letter

dated August 17,

1984).

The inspector

reviewed procedure

MP 12207. 1 dated

February

10,

1987,

Intermediate

Range Nuclear Instrumentation

Compensating

Voltage Adjustment,

and

found that the

necessary

changes

had

been

made.

Review of the licensee

's documentation

indicated that the procedure

has

not

had

problems

in this specific

area

since

the original violation.

Violation 250/83-41-01

and 251/83-40-01 is closed.

(Closed) Violation 250/84-29-01

and 251/84-30-02.

10 CFR 50.59 Evaluation

Not Made.

The original item in the inspection report discussed

Intake

and

Component

Cooling Water'(ICW/CCW) system

changes

which placed the

systems

outside of the Final Safety Analysis Report parameters

for which there

was

no safety evaluation

performed

(10 CFR 50.59).

The

NRC, letter of May 13,

e

1985

found the licensee's

written response

to the violation acceptable.

Subsequent

events of a similar nature

have occurred in the implementation

phase of this violation that have

been tracked

by the

NRC.

This violation

is

administratively

closed

and

tracked

under

Unresolved

Item

250,251/87-27-02.

Violation 250/84-29-01

and 251/84-30-02 is closed.

(Closed) Violation 250/84-29-02

and 251/84-30-03.

Technical Specification

Operability Not Shown.

The

NRC found the licensee's

written response

to

the violation acceptable

(May 13,

1985).

The

heat

exchanger

s associated

with

ICW/CCW

system still

demonstrate

degradation

problems

and

the

implementation

problems of this violation are being tracked

by the

NRC as

URI 250,251/87-27-02,

Based

on this tracking, the subject

1984 violation

is administratively

closed.

Sub-sections

(3)

and

(4)

under

the

1984

violation are acceptable

under

implementation.

Violation 250/84-29-02

and

251/84-30-03 is closed.

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)

4

(Closed)

IFI 250/83-41-02

and 251/83-40-02.

Failure to Implement Proper

Maintenance

and

Housekeeping

in Accordance

with guality Procedure

2. 12,

Revision 0.

At the time this item was open, oil soaked

lagging

from the

Unit 3 reactor

coolant

Pump

(RCP) lubrication oil system

had resulted

in a

fire.

The inspector

reviewed the following procedures:

AP 0103. 11

0-PME-061. 1

0-PME-041. 2

AP-090.73

Housekeeping

Reactor Coolant

Pump Oil Collection

System,

April 12,

1985.

RCP Motor Oil Fill and Drain, February

24,

1987.

equality

Control

Inspection

and Surveillance

Program,

March 10,

1987.

The

above

procedures

contained

adequate

instructions

to

prevent

the

reoccurrence

of the

above

mentioned fire.

The

PME procedures

have

been

reviewed

by the

Procedure

Upgrade

Program.

Additionally,

Maintenance

procedures

such

as

MP-2147.2

(Charging

Pump

Disassembly,

Repair

and

Assembly)

have

clean

up instructions

in their texts

and/or

places

for

initials in

the

text

as

a

means

of indicating

clean

up

has

been

accomplished.

IFI 250/83-41-02

and 251/83-40-02 is closed.

(Closed)

IFI 250/83-41-03

and 251/83-40-03.

Inadequate

pre-startup

valve

lineup for Safety Injection

and Containment

Spray Systems.

The cause of

this item was that the licensee

did, not

have

a formal valve lineup for

vent

and drain

valves within the

systems.

The inspector

reviewed

the

following procedures:

3-OP-062

4-OP-068

Safety Injection, June

18,

1987.

Containment

Spray System,

January

29,

1987.

.5

Both of the

system

lineup procedures

contain

valve 'alignment position,

check,

and

veri fication for vents

and

drains.

IFI 250/83-41-03

and

251/83-40-03 is closed.

(Closed)

IFI

250/83-41-04

and- 251/83-40-04.

Inconsistency

Between

AP-0103.5

and

OP-4103. 1.

conflicted

over whether

valve 837 was locked,

and if Unit 4 Safety Injection (SI)

pump suction

and

discharge

valves

had

the

correct prefix

on their identification tags.

Procedure

0-ADM-205 dated

July

18,

1987,

which

has

replaced

AP-0103.5

currently

shows

valve

837

as

locked

closed.

Procedure

34-OP-62

which

replaced

OP

4103. 1 currently indicates

the valve

as

locked closed.

At

present,

the Unit 4 SI

pump suction

and di scharge

valves

are correctly

labeled.

IFI 250/83-41-04

and 251/83-40-04 is closed.

(Closed)

IFI 250/83-41-05

and

251/83-40-05.

Fire Protection

Controls

During Welding.

At the time this item was open, during

a work evolution,

a firewatch did not

have

a

charged fire extinguisher.

The

inspector

reviewed the following documents:

MP-15537. 5

O-ADM-013.4

0-ADM-013

AP-0190. 67

Fire Protection

Equipment Surveillance of May 7, 1987.

Special

Interim Fire Watch Duties

and Training for

Appendix "R" Modification of June

9,

1987.

Fire Watch Requirements

and Duties of November 20,

1986.

Welding, Cutting, Grinding,

and

Open

Flame Work Safety

Procedure

of March 3,

1987.

The review of the

above procedures

indicated that there are at least

two

checks for extinguisher

conditions prior to work being

performed.

IFI

250/83-41-05

and 251/83-40-05 is closed,

(Closed)

IFI 250/84-23-15

and 251/84-24-15.

Evaluate

Licensee Ability to

Deal With Real

Time Procedure

Change

Requirements.

A real

time support

group which is

a

subgroup

under the

Procedure

Upgrade

Program

has

been

formed.

Additional personnel

have

been

hired to support

the real

time

procedure effort.

IFI 250/84-23-15

and 251/84-24-15 is closed.

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.

(Open). URI 250,251/86-18-13.

Licensee

to Provide

Loss

of

DC Procedure.

This item remains

open

as the last lice~see

action of NRC Bulletin 79-27,

which is discussed

later in this paragraph.

(Closed)

79-Bu-27,

Loss of Non-Class

IE Instrumentation

and Control

Power

System

Bus During Operation.

The inspector

reviewed

the

following

documentation:

Letter

Number

L-80-71

L-80-173

PTP-RE-85-124

JPE-PTRO-87-926

Date

March 3,

1980

June

6,

1980

July 9,

1985

May 20,

1987

Recipient

NRC

Site Licensing

Technical

Department

(site)

The

1980

FPL letters were the licensee's

response

to the bulletin.

The

inspectors

reviewed

the

following

NRC

IE Inspection

Reports

which

dealt with aspects

of the bulletin:

Number

a.

251/84-14

b.

250/84-29

and

251/84-30

c.

250,251/85-20

d.

250, 251/86-18

e.

250,251/87-07

250,251/87-10

g.

250,251/87"33

Item/Section

Item 02

Item 03

Item 04

Section

5

Section ll, Item 13

Section

2

Section

4.C

Section

5

Reports a,b,c,f,

and

g dealt with AC vital

bus portion of the bulletin.

Reports

d and

e above dealt with Loss of

DC Power.

Report

d (section ll,

item 13) which is

an unresolved

item that

has yet to be closed contains

the final

known action required of the licensee

under the bulletin 79-27.

For

Administrative

purposed

bulletin

79-27

is

closed.

URI

250,251/86-18-13

of

NRC Inspection

report

250,251/86-18 will track the

remaining action of the. bulletin.

(Closed)

84-Bu-02,

The'icensee's

response

to

IE Bulletin

No. 84-02,

"Failure of General

Electric Type

HFA Relays

in

Use in Class

1E Safety

Systems,"

was reviewed

and evaluated

during this inspection

period by the

Plant Systems

Section at Region II. Justification for closing-out this IE

Bulletin is

as

follows (Item

Number

cot respond

to action

items in the

bulletin):

.

la.

The licensee

stated

in his response,

dated July 20,

1984, that

he would replace

the relays identified in the bulletin as being

a potential

safety

problem.

Review of the

appropriate

work

orders

indicates

that

the

problem

relays

were

replaced

with

,qualified relays within the stipulated

time frame.

The licensee

confirmed that all

HFA relays mounted in the safety-related

4160

volt switchgear

were replaced.

We note here for the record that

the

HFA relay located in the compartment for the condensate

pump

is actually the bus clearing relay mentioned

in the licensee's

.

response.

1b and c.

The licensee

stated

in his response,

dated July 20,

1984, that

he did not have

any normally energized

HFA relays with nylon or

lexon

coil

spools

installed

in safety-related

applications.

Therefore,

the functional test

and visual

inspection

were not

required to be performed.

Also, it was not necessary

to provide

a basis for continued operation.

The stipulated

report

was

provided within the

required

time

period.

2 and

3

These

items were not applicable to Turkey Point.

In consideration

of the

above facts,

open

item 84-BU-02 is closed for

Units

3 and 4.

(Closed)

85-BU-02,

The licensee's

response

to

IE Bulletin

No. 85-02,

"Undervoltage

Trip Attachments

of Mestinghouse

DB-50 Type Reactor Trip

Breakers"

was reviewed

by Region II inspectors

according to the guidelines

in

Temporary

Instruction

2515/72.

Justification

for closing this

IE

Bulletin is

as

follows (Item

Number

correspond

to

action

items

in

the bulletin):

1.

The licensee

stated

in his response,

dated

December

9,

1985, that

he

performed

the required test

on the undervoltage trip device within

the specified time frame.

2.

The licensee

stated

in his response,

dated

December

9,

1985, that the

appropriate test procedure

was revised to include the conducting of a

force. margin test

on

the

undervoltage

trip devices.

Request

for

Procedure

Change,

OTSC No. 3734,

was reviewed.

This document revised

TOP206,

"Reactor Protection System-Periodic

Test (Unit 4 Only)", and

confirms the licensee

response

on this item.

3.

The licensee

stated

in his response,

dated

December

9,

1985, that the

specified

written instructions

were

issued.

Review of "Training

Brief ¹94" confirms the response.

4.

The required report was submitted within the stipulated

time period.

This bulletin

was

applicable

to only Unit 4.

In consideration

of the

above facts,

open item 85-BU-02 is closed for Units

3 and

4 and,

T2515/72

is closed for Unit 4.

7.

Onsite

Followup

and In-Office Review of Mritten Reports

Of Nonroutine

Events

(92700/92712)

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)

LER 250/85-035

Technical

Specification

Containment

Sump

Level

Indication.

The inspector

reviewed

documentation

for work performed

on

the

sump level indicators

LT-6308A and

B.

The documentation

included

the

component test sheets

and work orders.

LER 250/85-035 is closed.

Monthly and Annual Surveillance Observation

(61726/61700)

The

inspectors

witnessed/reviewed

portions

of

the

activities:

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.

following test

Nuclear Plant Operator

Logsheets,

4-0SP-201.3

Auxiliary Feedwater

System

Flowpath Verification, 4-0SP-075.5

Safety Injection

Pumps Inservice Test,

O-OSP-062.2

On June

27,

1987, with Unit 4 in Mode

5 and the

Reactor

Coolant System

(RCS) temperature

190F, the licensee

was making preparations

to heatup to

greater

than

200F.

TS require specific survei llances to be satisfactorily

performed prior to attaining

an

RCS temperature

of 200F.

Specifically

TS

4. 1, Table

4. 1-2,

Item 10,

requires that the'oron concentration

of each

of the three cold leg accumulators

be

sampled

and verified to be

1950

ppm

or greater prior to

RCS heat

up

above

200F.

The

4A and

4B accumulators

were satisfactorily

sampled

and

had boron concentrations

of 2215

ppm and

2240

ppm,

respectively.

The

4C

was

found to

be

empty,

therefore

a sample

could not be taken

and

compliance

to requirements

of

TS 4. 1,

Table 4. 1-2,.

Item 10

could

not,be

obtained.

Concurrently,

TS 3. 15. 1, Overpressure

Mitigation System

(OMS), requires

that with

RCS

pressure

boundary integrity established,

the valves

required to fill the

accumulators

must be closed with power removed, until the

RCS temperature

is greater

than

380F.

In order to comply

with the requirements

of TS

4. 1, Table 4. 1-2,

Item

10

and

TS 3. 15. 1 the licensee

should

have

placed

Unit 4 in

a lesser

condition of operation

by reducing

RCS temperature,

depressurizing

the

RCS and then filling the

4C accumulator.

Contrary to the

above,

the licensee

did not comply with TS .4.1,

Table

4'. 1-2,

item

10.

'The

required

sample

was not taken prior to e'xceeding

200F.

The licensee

delayed filling and sampling the

4C accumulator until

temperature

was increased

above

380F,

when

TS 3. 15. 1 restrictions

on valve

MOV-869 were

no

longer applicable.

The

NRC

was

not

informed of this

decision.

This discrepancy

was identified

by

NRC inspectors

during

routine

log

reviews

on July 23,

1987.

The licensee

was informed that the decision not

to implement

the

TS required surveillance

constituted

a violation of TS 4.1, Table 4.1-2,

Item 10. (251/87-35-01).

This violation is similar to violation 250/85-24-03,

issued

on July 30,

1985.

On

June

22,

1985

a

Unit

3 plant

heatup

was

in progress

in

accordance

with Operating

Procedure

(OP)

0202. 1, dated

April 12,

1985,

entitled

Reactor

Startup - Cold Conditions to Hot Shutdown

Conditions.

Section

3. 12'3 of the procedure

required that the boron concentration

in

each

be verified to be at least

1950

ppm prior to exceeding

an

RCS temperature

of 200F.

An on-the-spot-change

(OTSC)

was

approved

to

move this

requirement

to

another

section

of

OP

0202. 1 which

was

not

performed until after

the

required

sampling

temperature

of Technical

Specification

4. 1, Table

4'-2, item

10 was exceeded.

FPL responded

to this violation in letters

L-85-342, dated August 29,

1985

and L-86-31, dated January

24,

1986.

The licensee

stated

in part:

"At the

time of the

incident,

the

accumulators

were

drained

and

preparations

were being

made for a reactor coolant

system

heatup to

greater

than

200F.

'A conflict in

sampling criteria

vs.

equipment

availability criteria

was misinterpreted

thus allowing

OTSC to be

made to Operating

Procedure

(OP)

0202. 1,

"Reactor

Startup - Cold

Shutdown

to Hot

Shutdown

Conditions",

that

moved

the

sampling

to

a

later

step

in

the

procedure.

Another

factor

contributing

to

this

misinterpretation

was

the

Overpressure

Mitigating System

(OMS) Technical Specification which does not allow

opening of motor operated

valve (MOV)-869 with reacto~ coolant system

temperature

below

380F.

MOV-869

is

opened

when filling the

accumulators.

The

OTSC

was cancelled

in order to re-establish

the

TS requirement

back into the procedure

at the proper

sequence."

In both instances,

when

unable to obtain

a required

sample,

licensee

personnel

chose to omit; rather than meet,

the requirement.

The

decision

not to fill the

accumulators

was

influenced

by

a

licensee

interpretation

of TS 3. 15. 1

which precluded

the

opening of accumulator

fill valve MOV-869.

9.

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.

The following.items

were

considered

during this review,

as appropriate:

'hat

LCOs were met while components

or systems

were removed from service;

that

approvals

were

obtained prior to initiating work; that activities

10

were

accomplished

using

approved

procedures

and

were

inspected

as

applicable;

that procedures

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

gC records

were maintained; that activities were

accomplished

by qualified personnel;

that parts

and materials

used

were

properly certified; that radiological controls were properly implemented;

that

gC 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 approved

gA program;

and

that housekeeping

was actively pursued.

a.

Steam Generator

(SG) Feedwater

Regulating

Valve Concerns

On

July 28,

1987,

Unit

4

was

required

to

reduce

power

to

approximately

15% to facilitate repairs

to

the

4A

SG

regulating valve,

FCV-4-478.

The

two piece

coupling which secures

the valve actuator

to the valve

stem

was

observed

to

be shifting,

allowing valve

stem

movement

independent

of actuator

motion.

The

coupling for FCV-4-478 was replaced with a coupling from a Unit 3

SG

feedwater regulating valve.

NCR-87-187

was generated

to address

this

issue.

The

recommended

di sposition

of the

non-conformance

report

(NCR)

was to install

a bolt into

a

spare

connection

port in the

coupling to maintain proper orientation.

On August 5,

1987,

Unit 4

was

required

to

reduce

power again to approximately

15% due to

a

similar problem with the coupling of the

4C

SG feedwater

regulating

valve.

Troubleshooting

revealed that during the previous re-assembly

of the valves,

the coupling halves

for the

4B and

4C

SG feedwater

regulating

valves

had

been

interchanged:

Due to inexact machining

the

interchanged

coupling

valves

didn't

provide

full

thread

engagement,

allowing

independent

actuator

and

stem

motion.

The

couplings

were restored

to their appropriate

valves

and set

screws

were installed

and the unit was returned to

100% power.

An Event

Response

Team

(ERT),

(FPL correspondence

PTN-TECH-87-534),

was formed to address

the feedwater

regulating

valve problems.

The

long term corrective action is to replace

the existing couplings with

couplings

designed

by

Fisher

Valve,

during

the

of

sufficient duration.

b.

Feedwater

System Piping Thickness

Concerns

The licensee

established

an initial program to perform ultra sonic

wall thickness

inspections

on

selected

fittings in the

B Emergency Diesel Generators

system

in response

to the

Surry .pipe rupture

event

(IEN 86-106).

Engineering

correspondence

JPE-PTPM-87-616,

dated April 27,

1987,

delineated

the

inspection

program

and

acceptance

criteria.

The

acceptance

criteria was that the

measured

thickness

must

be greater

o

11

than

Tm

+ 0.075

inches,

where

Tm is the

minimum wall thickness

required to withstand internal pressure,

or an

NCR must

be generated.

The following NCRs were generated:

"87-0103

"87-0104

87-0108

"87-0110

87-0118

(C-0535-87)

87"0195

(C-0500-87

87-0198

C-0510"87

(C-0555-87)

"C-0541-87

"C-0542-87

C-0643-87

"C-0644-87

C-0647-87

The (*) indicates

NCRs which resulted

in at least

one fitting being

replaced.

More specific information and

a more formalized

long term

program will be

submitted to the

NRC in the licensee

response

to IE

Bulletin 87-01:

Thinning of Pipe

Walls in Nuclear

Power

Plants,

issued

July

9, 1987.

Licensee

responses

are

requested

within sixty

days from the receipt of the bulletin.

The

following

PWOs

and

procedures

were

reviewed

in detail

for

specific support to these

and other maintenance

activies:

PWO 6845 - Repair of 4-FCV-478

PWO 6844 - Repair of 4-FCV-488

PWO 6870 - Repair of 4-FCV-498

PWO 6146

Calibration of the

A

PWO 0263 - A EDG fuel oil filter

PWO 6152

A EDG skid tank level

0-GMI-102. 1, Troubleshooting

and

1987

¹ see violation 250,251/87-35-03

EDG air start pressure

indicator ¹

change

and strainer

cleaning ¹

indicator troubleshooting ¹

Repair Guidelines,

dated

March 17,

10.

Operational

Safety Verification (71707)

The inspectors

observed

control

room operations,

reviewed applicable logs,

conducted

discussions

with control

room

operators,

observed

shift

turnovers

and

confirm'ed 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.

By observation

and direct interviews,

verification was'ade

that the

physical security plan was being implemented.

Plant

housekeeping/cleanliness

conditions

radiological controls were observed.

and

implementation

of

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.

,

o

e

12

The

inspe'ctors

walked

down accessible

portions of the following safety

to verify operability and proper valve/switch alignment:

A and

Auxiliary Feedwater

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

0

Fire Watch

Found Asleep

on Duty

AP-15500,

entitled

Fire Protection

Program,

revi sion dated July 9,

1987,

section

9.4. 1 requires

in part that

backup

suppression

be

established

as

compensatory

action during fire protection

impairment

of automatic fire suppression

systems

such

as

the

halon

system.

Section 9.5.3

specifies,

in part,

that

the

posting

of continuous

firewatch is an acceptable

compensatory

action

when the halon

system

is

impaired.

TP-347,

entitled'C

Equipment

and

Inverter

Rooms

Supplemental

Cooling Monitoring and Standby Condition, revision dated

June

25,

1987,

section

5. 1. 1, requires

that anytime door

108A-1 is

maintained

open,

a continuous firewatch shall

be established

to close

the

door

(108A-1) within

60

seconds

of

sounding

of the

Halon

Activation Alarm.

On July 29,

1987, during

a routine tour performed

by a member of the

equality

Assurance

Department,

a

continuous fire watch

posted

in

accordance

with AP-15500

and

TP-347

was

found to

be

asleep.

The

continuous

fire watch

post

was

established

in early

June,

1987

because

fire door

108A-1

was required to

be

open

as

specified

in

evaluation

JPE-LR-87-020,

Revision 2, entitled'urkey

Point Units 3

and

4 Safety

Evaluation

and Justification

for Continued

Operation

With Loss of Heating, Ventilating,

and Air Conditioning

(HVAC) to

DC

Equipment

and Inverter

Rooms.

TP-347

was established

to implement

the requirement of evaluation

JPE-LR-87-020.

The .failure of the fire watch to remain

awake

and attentive to his

responsibilities

could have precluded

the effective operation of the

inverter room halon system.

The failure to have

a continuously alert

fire watch constitutes

an

inadequate

implementation

of AP-15500

and

TP-347

and is an example of violation (250,251/87-35-02).

Although this discrepancy

was identified by the licensee

and promptly

compensated

for,

a violation is

being

issued

because

of

past

occurrences

of

personnel

sleeping

on

duty.

Inspection

Reports

250,251/87-05

and 250,251/87-11

document-7

occasions

when

security

personnel

were

found to

be sleeping.

.The licensee

identified

6 of

the

7 examples.

Escalated

enforcement

action was taken

on April 21,

1987,

when

a Severity Level III violation was issued.

The fire watch

13

was

an

employee

of the

same

contractor

which

suppl ies

security

personnel

but he had

no security responsibilities.

However, his task

was important to the proper functioning of the halon fire suppression

system.

Discussions

with licensee

management

revealed

that this

problem is viewed seriously

by

FPL management.

Corrective actions

are being developed

and will be reviewed

subsequent

to the licensee's

response

to the Notice of Violation.

Valve Lock Found Incorrectly Attached

On

August 6,

1987,

during

a

routine plant tour,

the

inspectors

noticed that instrument air valve 3-40-856

was not locked as required

by procedure

O-ADM-205, Administrative Control of Valves,

Locks,

and

Switches,

revision dated July 18,

1987.

The valve

was closed,

as

required,

and the lock was closed,

However,

the lock wire was not

properly

engaged

through

the

valve

handwheel

and

around

the

instrument air pipe.

Consequently,

the valve could

be

opened

while

the

lock

remained

engaged.

The

licensee

promptly corrected

the

discrepancy.

TS 6.8. 1

requires,

in part,

that

procedures

be

established

and

implemented which meet or exceed

the recommendations

of Appendix A of

USNRC

Regulatory

Guide

(RG)

1.33.

RG 1.33,

Appendix A,

item 1.c

specifies

that procedures

should

be developed

to control

equipment

through locking and tagging.

The failure to maintain

valve 3-40-856

locked is an example of violation (250/87-35-02).

Emergency

Diesel Generator Operability, Test Requirement

On July 30,

1987,

at 4:45 p.m.,

the

A

EDG was

taken out of service

for routine preventive

maintenance.

Approximately twenty hours later

at 12:00

noon

on July 31,

1987,

the maintenance

was completed

and the

post maintenance

operability test run was

commenced.

Rated

speed

was

attained

but

the

"A"

EDG tripped

due to high crankcase

pressure.

Root cause troubleshooting

and analysis

were

commenced.

Operations

personnel

were

advised

by Regulatory

Co'mpliance

to

be

prepared

to run the

B EDG if the

A EDG was not returned

to service

by

4:00 p.m.

At

3:37 p.m.,

approximately

twenty

three

hours

after

being

taken

out of service,

the repairs to the A EDG were completed

and the post maintenance

operability test run was

recommended.

The A

EDG was successfully

run, the procedures

and accompanying

data

sheets

were reviewed

as satisfactory

and the

A

EDG was returned

to service

at 7:35 p.m., twenty six hours

and fifty minutes after being

removed

from service.

ll

TS 3.7.2.b

states

in part that

power operation

may continue if one

diesel

generator

is out of service

provided

the

remaining diesel

generator

is

tested

daily

and its

associated

engineered

safety

features

are operable.

Contrary

to

the

above,

the

licensee

failed to

comply with the

requirements

of TS 3.7.2.b,

in that

on July 30-31,

1987,

the

A

EOG

was

out of service

for greater

than

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

(26 hrs.,

50 min.)

without verifying the operability of the

B EDG.

This is

a violation

of TS 3.7.2.b (250,251/87-35-03).

The licensee

took .prompt corrective actions to prevent recurrence

of

this event.

FPL correspondence

PTN-TECH-87-539 was generated

stating

in part that, the more conservative

interim TS 3.8. l. 1, applicable to

EDG operability,

be

implemented

immediately.

Operations

personnel

were instructed

to meet the following guidance

in order to achieve

regulatory compliance.

(1) If an

EDG is discovered to be inoperable, test the remaining

EDG

for operability within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s of the discovery

and at least

once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter;

(2) If an

EDG is

to

be

taken

out

of service

for preplanned

maintenance

or other modifications, test

the

remaining

EDG for

operability

prior to taking

the

EOG out of

ser vice for the

maintenance

and test

the

remaining

EDG at least

once

per

24

hours thereafter.

This will ensure

that 'the units

are

not

placed

in

a position

where both

EOGs are out of service at the

same time.

Component

Cooling Water

(CCW) Heat Exchanger

Cleaning

Concerns

During

review of the

Plant

Supervisor-Nuclear

(PSN)

and Shift

Technical

Advisor

(STA)

logs

of July 23,

1987,

the

inspectors

identified an apparent

discrepancy

in the

CCW heat exchanger

cleaning

requireme'nts.

The

PSN log

and

equipment

out of service

(EOOS)

log

documented that the

4C

CCW heat exchanger

was declared

out of service

for

cleaning

at

8:00 p.m.

on

July 23,

1987.

The

ICW inlet

temperature/date

and

CCW heat exchangers

in service curve,

in the

PSN

log indicated that plant conditions existed

such that

a valve watch

operator

was required

to

be

posted

at

TCV-2201 prior to removing

a

heat

exchanger

from service.

The

STA log conflicts with the

PSN log

entries,

in that it documents

the

4C

CCW heat

exchanger

as

being

declared

out of service

and

a valve watch operator

posted at TCV-2201

at 10:00 p.m.

on July 23,

1987.

The inspectors'oncern

was that had

the

4C

CCW heat exchanger

been

removed

from service at 8:00 p.

m. and

the valve watch operator

not been

posted until 10:00 p.m. that for

this

2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period the licensee

would have

met the

requirements

of

safety

evaluation/JCO

JPE-6-85-38,

Revision 3.

More importantly in

this

scenario

the

Unit 4

CCW system

would not

have

been

able to

accept

and dissipate

the

MHA heat load of 12.0

x 10E6 BTU/HR.

The

inspectors

immediately

brought

this

concern

to

licensee

management,

who in tUrn initiated

an investigation.

Personnel

on

shift July 23,

1987 were contacted

and time

sheets

were

reviewed.

The licensee

concluded that the

STA log entries

were in error.

The

15

4C

CCW heat

exchanger

was in fact removed

from service at 8:00 p.m.

on July 23,

1987, but

a reactor

operator trainee was'roperly

posted

as the valve watch operator at TCV-2201 at 7:30

pm.

The inspectors

independently

reviewed the time sheets

of the trainee

and

interviewed

him to

ensure

that

he

in fact

had

been

properly

posted

and informed of his responsibilities

as the valve watch.

The

inspectors

concern that although

not documented

in quality records

that

. the

requirements

of

Safety

Evaluation/JCO

JPE-L-85-38,

Revision

3 were satisfied.

I

The

licensee

initiated corrective

actions

to

improve

programmatic

controls.

correspondence

PTN-TECH-87-462

delineated

,improved

measures

to ensure that

CCW heat

exchanger

design basis

heat

removal

capability is maintained throughout the cleaning process.

ll.

Engineered

Safety Features

Walkdown (71710).

The inspectors

performed

an inspection

designed

to verify the operability

of the Containment

Spray

(CS)

system

by performing

a complete

walkdown of

all

accessible

equipment.

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.

Valves were in proper position, breaker alignment

was correct,

power

was available,

and valves were locked/lockwired

as required.

e.

Local

and

remote

position

indication

was

compared

and

remote

instrumentation

was functional.

Breakers

and instrumentation

cabinets

were

inspected

to verify that

they were free of damage

and interference.

Conditions

that

were

noted

and

brought

to the attention

of licensee

include:

Unit 3

Valve 3-890A - boric acid on

a flange stud

Valve 3-891A boric acid

on the packing flange.-

PWO dated 5/5/87

Slight boric acid buildup on the

3B Containment

Spray

(CS)

pump

i

16

Containment isolation valve (CIV) tags not hung

on Unit 3

CS CIVs:

3-880'

and 3-890'

PWOs outstanding

on valves;

3-844B - since 6/23/86

3-880B

since 6/23/86

3-891B - since 5/5/87

Unit 4

Boric acid buildup on the

4A CS

pump

PWO dated 7/5/87

Clearance

order 4-86-8-053

hung

on valve 4-496T

12.

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.

On July 28,

1987, while Unit 4 was at

100% power, it was discovered that

the

4A Steam Generator

(SG) feedwater flow control valve actuator

was not

properly coupled to the valve

stem.

Automatic closure

on the valve

was

not affected,

but it was determined that the valve would not have remained

properly

seated

after

receipt

of

isolation.

In

order

to

faci lttate repairs,

a

power

reduction

to approximately

15%

power

was

required.

An

ERT

was

formed

in

response

to this

event.

Refer

to

paragraph

9.

On July 29,

1987, while Unit 4 was at

100% power,

a voltage spike occurred

in the process

radiation monitor system

(PRMS) rack and resulted

in

PRMS

4-R-11

tripping.

Containment

ventilation

isolation

and

control

room

ventilation isolation occurred

as

designed.

The voltage

spike occurred

when

an

I&C technician .increased

the

PRMS

R-15 high voltage

supply from

1000 volts to 1500 volts.

PRMS R-11 was reset.

On July 31,

1987, while Unit 3 was in Mode 5, the reactor trip breakers

opened

due to Nuclear Instrumentation

System

(NIS) channel

N-36 (spiking

high) high flux trip.

Trip breakers

were reset

and tripped manually to

verify SOE printout received.

N-36 high flux trip did not printout due to

the

channel

being inhibited.

N-36 was placed

in level trip bypass

and

taken

out

of

service

under

clearance

3-87-7-156.

A

suspected

bad

cable/connector

was the probable

cause of trip.

17

13.

On July 31,

1987, while Unit 3 was in Mode 5,

a class

B fire occurred at

1:20

pm,

east

of Unit 3 Turbine Plant Cooling Water

(TPCW) pumps,

due to

a hydrogen

gas

leak

near

a construction activity involving grinding

and

welding.

The leak- was identified on

a fitting between drain valve 3-4617U

and PI-3-1059E.

It was located

under

a construction

work area,

covered

by

hot work permit ¹7-528, for Control Work Order

(CWO) A-433.

The fire team

responded

and using

C02 extinguishers

put out the fire.

A water hose

was

also

used to cooldown the piping.

The hydrogen

supply from the gas

house

was

isolated.

The

leak

was

repaired

by

1&C

and

checks

for other

leaks-identified

and repaired.

Fire was reported out at 1:30

pm.

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

NRC 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'eal

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.

- - 14.

Design,

Design

Changes,

and Modifications

(37700)

An

inspection

was

conducted

to ascertain

the

licensee's

methods

of

assuring

that

design

changes

and

modifications

meet

the

review

and

approval criteria specified

in the Technical

Specifications

and

10 CFR 50.59.

of design

changes

were

documented

in

NRC

Inspection

Report 250,251/87-14,

paragraphs

8a through 8c.

In

a

May 21,

1987 letter the licensee

was informed that the emergency diesel

generator

sequencer

wiring discrepancies

described

in paragraphs

8a

and

8b of the

report were

under consideration

for escalated

enforcement

action.

In

a

July 21,

1987 letter

the

NRC concluded

that

the wiring discrepancies

represented

a Severity

Level IV violation.

The

sequencer

wiring errors

were

found because

the licensee

conducted, an

expanded wiring evaluation

.

after identifying

and correcting

a protection

relay discrepancy.

The

sequencer

wiring discrepancies

were

licensee

identified,

were

promptly

corrected

and could not reasonably

be expected

to have

been

prevented

by

the corrective action'rom

a previous violation.

Consequently,

pursuant

to the provisions of 10 CFR Part 2, Appendix C,

no notice of violation was

issued for the item.

18

Additional Plant

Change Modifications (PC/Ms) were evaluated

in the areas

of

containment

spray,

auxiliary