ML17354A597

S stem

Ins ection Sco

e

71707

and 61726

The inspector. performed

a Unit 3 and Unit 4

RHR system'walkdown,

and

observed

.the system engineer

perform the monthly flow path verification ..

surveillance

on the, Unit 4

RHR system.

b.

Observations

and Findin s

The

RHR surveillance

was described

in procedure

4-OSP-202. 1,. "Safety

Injection/Residual

Heat Removal

Flow Path Veri.fication .'he

surveillance verified the system flow path

and power avai labi.lity to the

required

components

at the corresponding

operating

mode:.

At the time of

~

02.4

the surveillance, both'nits were at

100K power.

The inspector

reviewed

the piping and instrument drawings

(P8 ID) for the

RHR system

and

verified that the surveillance

procedure

was consistent with the P8ID

flow path

and component electrical

requi rements.

The inspector verified

the correct

RHR valve positions

and electrical

requirements

in the

control

room panels.

Additionally, the inspector walked down the Unit 4

RHR system with the

system engineer

and observed the system engineer

perform the monthly

surveillance.

The inspector noted that two motor operated

valves,

(MOVs)

MOV 4-861B and

MOV 4-752B,

had traces of boron by the stem and

bonnet area.

The system engineer

noted that theses

valves

had already

been tagged with a

PWO for maintenance

work to be performed at the

upcoming outage.

The system engineer

provided significant detail

regarding the system flow path, operation,

history of the

RHR system,

and construction

and operation of the major components

in the system.

The inspector

noted the system engineer's

knowledge of the

RHR system to

be

a strength.

The inspector verified that the surveillance

schedule

for this test

was met and also reviewed the completed surveillance

packages

for the last six months

and verified proper licensee

reviews

and approvals.

Conclusions

At the completion of the surveillance,

the inspector concluded that both

units'HR systems,

were correctly aligned for standby operation.

The

systems

engineer's

knowledge of the

RHR system

was

a strength.

Control

Room Annunciator Status

71707

Numerous times during the inspection period. the inspectors

observed

that no control

room annunciators

were in an alarmed condition.

This

"blackboard" condition was noted

on different days

and shifts during the

inspection period.

The licensee tracks these off-normal (lit)

annunciators

in the daily plan-of-the-day

(POD) report.

Operations

. personnel

report at the morning

POD meeting

as to the number of lit

annunciators

and their status for resolution.

The inspector noted excellent support for operations

by both maintenance

and engineering in resolving lit (alarmed)

annunciator

issues.

Achieving and maintaining

a control

room "blackboard" condition for both

units demonstrated

strong operational

performance.

05

Operator Training and Qualification

05.:1

Simula'tor Trainin

71707 '.

.During the inspection peri'od, the'inspectors

reviewed

and 'assessed

the

-'icensee's

training department relative to simulator training for

operators,

and simulator

use

and consistency for observed plant p'roblems

and transients.

This included the following activities:

06

06.1

Unit 4 automatic trip on April 23,

1997,

(Reference

NRC Inspection

Report

Nos. 50-250,251/97-04

section 04.1),

Auxiliary Building Ventilation response

(section 02.1),

Loss of the

3C

NCC (section Ol.l), and

Routine licensed operator

simulator refresher training.

In addition

~ the inspector

reviewed

a new training device developed to

reinforce the self-checking practices of Stop-Think-Act-Review (STAR).

A "STAR Simulator" was developed

in-house with guidance

from the

industry.

This simulator replicates

a control panel with switches,

controls,

and indications.

The simulator hardware is used in conjunc-

tion, with an "operating procedure" to test

an individuals self-checking

and therefore

STAR efficiency.

Since this training device is new, its

overall impact has not been yet assessed.

However, this device appeared

to be

a very good training aid.

In conclusions

the operations training department

was proactive

and very

responsive

in assisting plant operations,

and provided excellent support

for operations.

In addition,

assurance

of simulator quality and

consistency with the plant were noted.

Operations

Organization

and Administration

Hang ement Staffin

and Availabilit

71707

On June 3,

1997, during the 7:15 a.m.

morning management

meeting,

the

inspector noted that the Plant

Manage

and his three direct reports

(Operations,

Maintenance,

and Work Control Managers)

and the Services

Manager were all absent.

The Operations

Supervisor

was present

providing management

coverage.

In addition, the Site Vice President

was

onsite.

The inspector

reviewed licensee

procedures

and

NRC requirements

relative

to management

staffing and availability.

TS 6. 1. 1 requires that the

Plant Manager designate

in writing the lines of succession

in his

absence.

The inspector verified that this was done.

The lines of

succession

were the Operations

Manager,

the Services

Manager,

and the

Operations

Supervisor.

The inspector also verified that the Emergency

P'lan staffing requi rements

were met.

The inspector also discussed this

issue with the Site Vice'President:

FPL policy- requi res either the Site

. Vice President

or the Plant"Manager

be available (onsite).

The

.

. inspector

was also. informed that* bo'th the, Plant and Maintenance

Managers

were .attending

a local industry conference

and .were available within

about

60 minutes..

The inspector concluded

that"management

staffing and availability were

appropriate.

08

Miseel 1 aneous

Oper ations Issues

08.1

Closed

VIO 50-250 251/96-13-02

92901

The violation resulted

when

a senior nuclear plant operator

(SNPO)

failed to follow the liquid radwaste

operating procedure

(OP).

Licensee

corrective actions

were detailed in a letter (L-97-043) dated

February 28,

1997.

These actions

included the following:

The

SNPO was disciplined

and

removed from operations,

A root cause analysis

was performed,

Shift supervision monitored radwaste operations for a two month

periods

Operations

supervision

discussed

the event with SNPOs

and other

non-licensed

operators,

The related

OP was reviewed

and revised

as appropriate,

SNPOs provided constant monitoring of any radwaste building

evolutions,

and

A remote alarm was added to provide the control

room with radwaste

system or building abnormalities.

The inspector

reviewed the licensee's

response,

verified corrective

actions,

and observed

selected

radwaste building evolutions.

Based

on

satisfactory observations

and inspector verifications'he violation was

closed.

II.

Maintenance

Ml

Conduct of Maintenance

H1.1

General

Comments

a.

Ins ection Sco

e

Maintenance

and surveillance

test activities were witnessed

or reviewed.

b." Observations

and Findin s

The inspector

witnessed

or 'reviewed portions.of"the,following mainte-

nance-activities

-in progress.

..3C. NCC repairs

(section 01. 1)

A AFW outage (section H1.2)

3A Sequencer

relay replacements

per procedure

O-PMI-024.4 Emergen-

cy Load Sequencer

Relay Replacement

and Inspection (section M1.3).

4A CCW pump overhaul

(section Ml.4).

3B

RHR pump overhaul

(section Ml.6).

The inspectors

witnessed or reviewed portions of the f'ollowing

surveillance

test

and inser vice test

( IST) activities:

AFW Train

1 testing (section Ml.2).

3A Sequencer

testing per procedure

3-0SP-024.2,

Emergency

Bus Load

Sequencer

Manual Test (section M1.3).

Procedures

3/4-0SP-201.2,

SI/RHR Flow Path Verification (section

02.3).

Procedures

3/4-075.5,

AFW System Flowpath Verification (section

M1.2) .

Procedure

3-0SP-019.1.

Intake Cooling Water IST (section M1.7).

Procedure

4-0SP-068.2,

Containment

Spray

IST (section M1.8).

For those maintenance

and surveillance activities observed

or reviewed,

the inspectors

determined that the activities were conducted in a

satisfactory

manner

and that the work was properly performed in

accordance

with approved maintenance

work orders.

The inspectors

also determined that the above testing activities were

performed in a satisfactory

manner

and met the requi rements of the

technical specifications.

c.

Conclusions

Observed

maintenance

and surveillance test activities were well

performed.

M1.2

A Auxiliar

Feedwater

AFW

Pum

and Train

1 Maintenance

a

b.

Ins ection.Sco

e

61726

and

62707

'I

The inspector. observed.

AFW system

main'tenan'ce

and testing.

.Observations

and Findi'n s

The licensee

removed the A.AFW pump and Train

1 from service for both

units in order to perform corrective.and

preventive maintenance

(PM),

and to conduct modification work (section

E8. 1).

The system

was

removed

'rom

service at 3:27 a.m.

on June 2, 1997.'elated

work included

Ml.3

electrical

speed

sensor

work, valve repacking,

pump

PMs, drain line

PC/M,

18C calibrations, air check valve replacements,

and other

miscellaneous

work items.

The work was completed,

post maintenance

and

surveillance

tests

were completed,

and the A AFW was declared

operable

at 6:20 a.m.

on June

3 for Unit 3 and at 7:00 p.m.

on June

3 for Unit 4.

The Unit 4 testing per procedure

4-0SP-075.6,

AFW Nitrogen Backup Test.

initially failed due to excessive

leakage through the air check valves

that had been replaced.

Metal shavings

were found when the replacement

check valves were opened

and inspected.

CR No.97-937 was written to

address this apparent

FME issue.

A metallurgical report concluded that

these

shavings

were from the manufacturing

process.

Subsequent

replace-

ments

and testing were completed successfully.

The inspector

reviewed the work packages

and clearances,

observed

work

in the field, and reviewed testing results.

The inspector

noted very

good involvement by system engineering,

strong maintenance field

oversight,

and positive operations

involvement and oversight.

Management, provided good coordination

and expectations

requirements.

Conclusions

In all. this safety

and risk related

equipment

outage

was well planned

and executed.

A safety system walkdown and surveillance test verifica-

tion determined the

AFW system to be appropriately aligned after the

maintenance.

During the walkdown, the inspector

noted poor preservation

of the Unit 4 train two flow control valves

and piping.

This was

discussed

with management.

The licensee appropriately

addressed

this

during the inspection period.

Unit 3 3A Se uencer

Issues

and Testin

61726

and 62707

On June 4,

1997. during routine monthly sequencer

testing

(procedure

3-

OSP-024.2,

Emergency

Bus Load Sequencers

Manual Test),

abnormal light

indications were observed

by the

RCO performing the test.

The 127X2/3A4

bus stripping relay red actuation light failed to energize

as expected

during step

7. 16 of the procedure.

Subsequent

testing. noted the light

(and therefore the relay) to be sluggish

and then function correctly.

CR No.97-940 was initiated,

system engineering

and

I8C personnel

were

notified,

and preparations

for relay replacement activities were made.

The licensee's

IKC group replaced

two suspect

relays per procedure

O-PMI-024.4.

Operations successfully

re-performed

procedure

.3-OSP-

,024.2..

The licensee

intends .to perform'a fai lure analysis. on the.,two

removed sequencer. relays.

E

The inspector

observed the maintenance

arid testing activities, verified

technical specification

compliance..

reviewed. the'R and procedures

used.

in the field, and discussed

these

issues with operations,.engineering,

and maintenance

personnel.

The inspector

noted conservatism,

excellent

.

. teamwork,

good procedure

use,

and strong over sight of field activi.ties.

Senior plant managemerit

were observed to be in the field and involved.

b

10

M1.4

Unit 4 4A CCW Pum

Overhaul

62707

As discussed

in NRC Inspection Report

No. 50-250,251/97-04,

the 4A CCW

pump was experiencing higher than normal

(but acceptable)

pump bearing

vibration.

The licensee

removed the

pump from service for overhaul

on

June 3,

1997.

When the bearing

was

removed.

the ball bearing retainer

ring was noted to be damaged.

The entire bearing assembly

was sent to

the metallurgical lab for analysis.

The

pu'mp was overhauled

per

procedure

O-CMM-030.3,

CCW Pump Overhaul.

Restoration

and retest

activities were satisfactorily completed

on June 5,

1997.

The inspector observed field activities and discussed

the work with

maintenance

personnel.

The inspector concluded that the overhaul

was

performed in a timely and positive manner

.

M1.5

Post Accident

H dro en Monitor

PAHN

Issue

a.

Ins ection Sco

e

61726 and 62707

The inspector

reviewed issues relative to one train of the Unit 4

PAHN

system which was found out-of-service.

b.

Observations

and Findin s

Ouring monthly surveillance

testing

on June 6,

1997, per procedure

4-0SP-094.2,

Hydrogen Monitoring System Flowpath Verification, the 48

PAHN inlet and outlet valves were found closed

by the system engineer.

The valves are internal to the

PAHN cabinet,

are not labelled nor

locked,

and are not on the system

P8ID.

However, these valves

do affect

the system flowpath and the valves being closed

caused the 48

PAHN to be

inoperable.

The licensee

entered

TSAS 3.6.5.a

(30 days).

re-opened

the

valves,

and initiated an investigation per

CR No.97-957 by an Event

Response

Team (ERT).

The

ERT concluded that the valves were inadvertently c1osed during 48

PAHN calibration

and troubleshooting

per procedures

4-PNI-094.2,

Containment

Post Accident Hydrogen Monitor Instrumentation

Channels

AE-4-6307A/8 Calibration,

and 0-GMI-102. 1,

I8C Troubleshooting

on

June 4,

1997.

The 48

PAHM was

removed from service

on June

2,

1997,

and

returned

on June 5,

1997.

The inoper abi lity period was 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months after

the 48

PAHN, return to service.

and 4 days from initial start of the

ca1ibration (e.g.,

when the component

was initially removed from

service)..

The redundant

4A PAHM and both Unit 3

PAHMs were verified to

be operable

and appropriately aligned.'hus.

no TSASs were violated.=

~-

The 'ERT concluded root cause to .be personnel

error

by an IK technician

who failed to follow the procedural

requii ements,of the. calibration and

'roubleshooting

procedures.

The I8C technician did not document the

valve positioning

as required.

The technician subsequently

forgot to

reopen the valves.

Further post, maintenance

checks did not check the

valve lineup.

Corrective actions included personnel

counselling,

procedure

enhancements,

personnel

retraining,

a check of all other

11

accessible

instrument valves with no operability issues

noted,

QA review

of the issue,

labelling and locking of internal cabinet valves,

and

event promulgation to site personnel.

In addition, the licensee is

enhancing their program for I8C valve manipulation to include valve

tagging

and supervisory final checks.

The inspector

reviewed the event, including the licensee's

investigation

and associated

documentation.

The inspector

independently verified all

PAHH systems.

including the redundant Unit 4 4A train.

TSAS and

surveillance

requirements

and procedures

were reviewed,

and were also

verified to be adequate.

No TS violations were identified.

However,

the fai lure to follow the troubleshooting

and calibration procedures

was

identified as

a violation.

This non-repetitive.

licensee-identified

and

corrected violation is being treated

as

a Non-Cited Violation (NCV)

consistent with Section VII.B.1 of the

NRC Enforcement Policy.

50-250,251/97-06-01,

Failure to Follow I8C Surveillance

was closed.

Conclusions

One

NCV for failure to follow procedures

by an

I&C technician

was

identified.

3B Residual

Heat

Removal

Pum

Overhaul

61726

and 62707

The licensee

overhauled the 3A RHR pump per

procedure

O-GMH-050.5,

Pump Refurbishment.

The pump had an increasing

seal leakoff flow the

past

few in-service tests.

The pump was

removed from service at 1:00

p.m.

on June

11,

1997.

The pump and motor train was moved to the

radwaste building to perform the wor'k.

A refurbished

motor and

a new

seal

package

were assembled

with the existing

pump impeller.

The

licensee

worked through delays associated

with the impeller reassembly.

The pump was returned to service at 8:57 p.m.

on June

13.

1997.

The inspector verified TSAS compliance,

reviewed the clearance,

monitored radiological aspects

of the job. walked down the redundant

3A train, observed portions of maintenance

and testing activities,

verified procedure

compliance.

and

IST results

and discussed

the work

with operations,

engineers,

HP, maintenance,

and

QA/QC personnel.

Very

good coordination

was noted.

The original schedule

was for 46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months ,

however the work was completed in about

56 hours6.481481e-4 days 0.0156 hours 9.259259e-5 weeks 2.1308e-5 months .

The inspector concluded that although

some delays were encountered,

the

licensee appropriately

planned

and executed this overhaul activity for

the

3B RHR pump..

Teamwork was noted .to be very strong.

Unit 3 Intake Coolin

Mater

ICM

Low'Flow Rate

Ins ection Sco

e

61726

On June

17

1997, during the quarterly inservice test

(IST) of the.3A

ICW pump. the

ANPS requested that 'the test

be stopped'ue

to a low flow

rate indicatjon through the operable

ICW header.

12

b.

Observations

and Findin s

Procedure

number 3-OSP-019. 1, Revision dated 5/22/97,

Intake Cooling

Water Inservice Test, described

the surveillance

test

requirements

for

the Unit 3 IST of the

ICW pumps

and valves.

At the time of the test,

Unit 3 was at

100K power,

B had been. declared

inoperable,

and the

Unit was in a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement.

Header

A was the operable

and

pump 3A was being tested.

The procedure

required that the two

ICW headers

be separated

from each

other prior to testing any of the three

ICW pumps.

Header separation

had been accomplished

by manipulating the following valves:

closing valve 3-50-307

and opening valve 3-50-309 located at the

intake canal

on the

ICW pump discharge

header;

closing valve POV-3-4883 (closed

from the control

room);

and

closing valve 3-50-350

and opening valve 3-50-340 located in the

CCW heat exchanger

room on the

ICW to the

CCW heat

exchanger

header.

ICW Pumps

3A and

3B were lined up to the operable

A.

Pump 3A was

operating

and

pump 3B was off but was operable.

Additionally,

ICW flow

was going through the

3A and

3B

CCW heat exchangers

from header

A.

Although header

B had been declared

inoperable.

pump

3C was operating

(on header

B) and there

was

ICW flow from header

B going to the Turbine

Plant Cooling Water heat,exchangers

and to the

3C

CCW heat exchanger.

The test

was being performed by two people.

One senior nuclear plant

operator

(SNPO)

was at the

CCW heat exchanger

room and the watch

engineer

was at the intake canal.

There was also

a senior

reactor

operator

(SRO) trainee

who was observing the test.

The trainee arrived

after the test

had been started

and was at the intake canal with the

watch engineer

.

Communication

among the

SNPO, the watch engineer,

and

with the control

room was via walkie talkie.

To obtain the required test flow rate through the 3A pump,

procedure

step

7. 1.26 requi red

an iterative throttling process

on valves located

in the

CCW heat exchanger

room,

namely,

3-50-380,

3-50-370,

and 3-50-

360.

The requirement

was to: 1) obtain flow separation

between the two

ICW headers,

which would be confirmed by getting

a zero differential

pressure

between the headers.

and

2) obtaining

a 1,540

GPM total

ICW

f'low through the.3A.and

3B

CCW heat exchangers.

with .neither heat

exchanger

going .above 10.000'gallons

pei

minute"(GPH).." Also, there was

a note preceding this step which described that the

ICW flow through the

Turbine'lant Cooling Water heat'"exchangers

may have'o

be throttled to

achieve the 1,540

GPN requi rement.

During the iterative 'throttling process

the

SNPO realized that the flow

rate in the operable

ICW header

had been throttled closed too much.

Consequently...there

'was

a low ICW flow rate in the operable

header.

13

During this evolution,

communications

via the walkie talkies with the

watch engineer

were not clear, i.e., too much radio static,

which caused

miscommunications with instructions

on which valves

had to be throttled.

Operator

logs show that there

was approximately 3,500

GPM flow through

the operable

header for

a short period (less than

5 minutes).

Having

realized there

was low flow through the header

and not being able to

communicate clearly with the watch engineer,

the

SNPO stopped all test

activities.

The

SNPO phoned the control

room and informed the

ANPS of

the low flow through the operable

header.

The Assistant

Nuclear Plant

Supervisor

(ANPS) requested

that the valves

on the operable

be

opened

and that the test

be stopped.

Condition Report

No.97-994 was

initiated.

Engineering

performed

an operability assessment

and determined that the

ICW/CCW systems

had remained within the design

bases

during the short

deviation

ICW low flow condition.

Heat exchanger

CCW outlet temperature

data indicated that the temperature

during the low flow condition

increased

a few degrees

and reached

approximately

96 degrees

Fahrenheit.

versus

the

150 Degrees

Fahrenheit

design

bases

temperature limit.

Additionally, the licensee

refer red to two safety evaluations that had

been performed

on this system to confirm their operability assessment.

The licensee

concluded the reason for having obtained the low flow

condition was because

the IST procedure did not provide sufficient

detai 1 during the iterative throttling process

and allowed for misinter-

pretations of the requirements.

In addition, there

had been

miscommuni-

cation during the throttling process

which led to the wrong valve being

throttled.

Significant changes to the procedure

(Revision dated 6/19/97)

and other

.

improvements

included the following:

Explicitly describing the throttling process

requi rements.

Specifically, step 7.1.26 was described in more detail,

and

additional

steps

were added to exclude having to interpret the

, throttling process

requi rements.

. The IST now requires

one

SRO and three operators to perform the

surveillance.

Two operators

are stationed at the

CCW heat ex-

changer

room and one operator

is stationed at the intake canal.

The

SRO is in charge of the test,

and maintains contact with the

operators

and with the control

room at all times.

~

Marked-up

P8 IDs are to. be reviewed'at

the- pre-.job briefing by

" everyone performing the IST.

4

1

I

~ -', Three point radio 'communication 'is to be used at all. times

and the

'communication

system is to be checked to assure it is working

properly.

The inspector

reviewed the

CR, logs'he

above mentioned safety evalua-

tions,

UFSAR section 9.3.2,

and discussed this event with operators,

'

14

engineers,

and management.

Corrective actions were verified.

Operators

involved in test were also interviewed.

The inspector attended

the pre-job briefing for the subsequent

ICW 3B

IST surveillance

and observed the surveillance

being performed.

This

surveillance

was performed using the

new procedure

and guidelines.

Overall the inspector

found that the surveillance

was well performed.

The inspector

noted that the system engineer

was at the briefing and

observed the surveillance

and was available

for questions

throughout the

test.

The pre-job briefing was very thorough.

Marked up

P8 IDs were

reviewed

and the

SRO went over everyone's

part of the surveillance.

Changes

in the procedure,

system alignments

and implications to plant

operations

were discussed.

Communication practices,

self'-checking

(STAR), and personnel

safety were also reviewed.

During the IST, there

were two operators

at the

CCW room and one operator at the intake.

Communications with the operators

and with the control

room was good,

and the roving SRO continuously verified activities in the

CCW heat

exchanger

room and at the intake canal.

However, the inspector

noted

that during step 7.2.21 'he operators

at the

CCW heat exchanger

room

did not unlock valve 3-50-370.

Step 7.2. 21 requi red unlocking valves 3-

50-380

and 3-50-370 in preparation for the throttling process to obtain

the required flow through the

3B pump.

The inspector

asked the opera-

tors why they had not unlocked the 3-50-370 valve.

The operators

replied that .ICW flow through the 3B heat exchanger

was sufficient and

they believed the valve was not going to have to be throttled.

Although

through the completion of the test valve 3-50-370 was not throttled, the

inspector

noted that the roving SRO was not informed that the valve was

never unlocked until after the throttling process

had been completed.

The licensee

corrected this minor issue on-the-spot

and is considering

future procedure

enhancements

c.

Conclusions

Through interviews,

procedure

and log reviews,

P810 reviews

and observa-

tion of the

ICW 3B pump IST surveillance,

the inspector confirmed the

licensee findings and concluded the following:

~

The

SNPO acted conservatively

and,demonstrated

a strong safety

focus when stopping the test

and calling the control

room to ask

for help.

~

Engineering, support to operations

and maintenance

and teamwork in

resolving this issue

was noted

as a-.strength.

~

Clarification and improvement

made to the procedur'e resulted in'

good

ICW 3B pump IST test performance.

15

M1.8

Containment

S

r a

Pum

Inservice Test

IST

Ins ection Sco

e

61726

The inspector

observed

the Unit 4 Containment

Spray System Inservice

Test and reviewed the applicable procedure

no. 4-0SP-068.2.

Observations

and findin s

The inspector

noted that in calculating the differential pressure

across

the pump, the procedure

used the static suction pressure

instead of the

dynamic suction pressure,

i.e.. the static suction pressure

was

subtracted

from the dynamic discharge

pressure.

Further,

during the

testing the inspector

observed that the suction pressure

increased

when

the

pump was operating.

This increase

in suction pressure

was observed

on both the 4A and 48 containment

spray

pumps.

These observations

were later discussed

with the system engineer.

He

explained that

a number of years

back the procedure

had been

changed.

The system engineer

noted that the suction pressure

tap was in a very

turbulent, area

between

elbows.

The reason for the procedure

change

was

because

engineering did not believe that the increase in pressure

on the

suction side during pump operation

was

a correct reading,

and

consequently

chose not to use it in calculating the

pump differential

pressure.

The inspector

requested

to see the documentation

describing

the procedure

change

and any engineering analysis/testing

that was used

to support the change.

The licensee is gathering the information.

The

inspector intends to review this issue in the next inspection report.

The inspector noted during the vibration readings that the position

labels for the accelerometer

were not installed in the 4A containment

spray

pump.

This was later discussed

with the IST coordinator

and he

indicated that the

pump had been refurbished

and that they were planning

on putting the position labels

on the pump.

The inspector

verified the measuring

& test equipment

(M&TE) calibration

data

on the instruments

used for this surveillance.

However, the

inspector

noted that the accelerometers

.did not have any calibration

data.

In later discussions

with the

M&TE supervisor,

the licensee

indicated that the accelerometers

were not calibrated but they were

checked for hard failures per the vendors

recommendations.

The licensee

is reviewing enhancement

to the calibration program and including

obtaining .a shaker table -to verify calibration data

on the

accelerometers.

. " c... Conc'lusions ..

The containment

spray testing

was well performed.

The licensee

was

responsive to the inspector's

qu'estions.

16

H2

Maintenance

and Materia Condition of Fac'ilities

and Equipment

M2.1

Unit 4 Power 0 crated Relief Valves

PORVs

a.

Ins ection Sco

e

62707

The inspector

reviewed the status of the Unit 4 PORVs.

b.

Observations

and Findin s

Following the Unit 4 restart

from the April 23,

1997, automatic trip,

Pressurizer

Relief Tank

(PRT) parameters

were noted to be-increasing.

This included

PRT pressure,

levels

and temperature.

and

PORV downstream

tail pipe temperature.

The licensee initiated

CR No.97-830,

a problem

status

summary,

and

PWOs for each

PORV.

The licensee's

system engineering

group initiated an investigation

and

performed the following actions:

Checked for other

PRT in-leakage

sources

and none were found,

Measured

PORV block valve stem packing leakoff and none was found,

Quenched

and drained the

PRT,

Isolated

each

PORV (one-at-a-time)

and measured

leak rates,

Planned to work each

PORV during the September

1997 outage,

Continually monitored

PRT parameters,

Continued operation with one

PORV (PCV-4-456) isolated (e.g.,

block valve MOV-4-535 closed),

Noted an identified leak rate of 0.03

gpm with one

PORV isolated.

and

Issued

a problem status

summary by the system engineer.

The inspector

reviewed the above mentioned documentation,

and

TS 3.4.4

for the

PORVs and their block valves.

The TS compliance

was

appropriate,

and oper ation with one block valve closed

was addressed

on

the operator work-around listing.

The inspector noted very good

engineering

support of this Unit 4 issue".

Operator. awareness

of these

issues

was also very good:

The inspector

intends to.follow 'outage

repair activities and root cause deter'mination during the upcoming Unit

4 refueling outage.-

0

Conclus ions

17

H2.2

M2.3

The licensee appropriately

responded to the leaking Unit 4 PORVs.

Unit 4 Hain Steam Safet

Valve

HSSV

62707

During

a routine plant tour, the inspector

noted

an unusual

noise

emanating

from one of the Unit 4 HSSVs.

RV-4-1405 had

a rattling and

"ringing" noise coming from the valve top works which was the spindle

knocking against the compression

screw.

The inspector

pursued this issue

and noted than

a

CR (No.97-821)

and

a

PWO had been initiated,

as well

as

a problem status

summary.

The vendor (Dresser)

had been contacted

and an operability assessment

had been performed.

The licensee

concluded that the condition was from harmonic vibration of

the valve spindle against the compression

screw.

The tuned frequency of

the valve spindle was probably altered

when the valve lifted during the

April 23,

1997, Unit 4 trip (NRC Inspection

Report

Nos. 50-250,251/97-

04).

With vendor input, the licensee

concluded that the valve was

operable

per TS 3.7. 1. 1 and that the liftsetpoint

was unaffected.

Corrective actions

included operator-rounds

monitoring of the "ringing",

bi-weekly check for seat

leakage,

operations briefings per the status

summary document,

and repairs

planned during the September

1997 outage.

In addition,

longer term corrective actions including the use of shims

to prevent the internal

components

from contacting

each other are being

pursued.

The inspector

reviewed the operability assessment

in the

CR and related

corrective actions.

The inspector periodically toured the Unit 4 steam

platform and monitored this and other

MSSVs.

The inspector

concluded

that system engineering

support of this maintenance

issue

and the

operability

assessment

were appropriate.

Operator

awareness

was also

appropriate.

Reactor Coolant

Pum

RCP

Status

Ins ection Sco

e

62707

The inspector

reviewed several

issues

associated

with the

RCPs.

Observation

and

a Findin s

CYCLE 14

. CYCLE '15

9/5

.

0/0

12/1

0/0

'/0

The inspector'eviewed

the Unit 3 RCP motor oil consumption

over the

. past

few operating cycles. (h.g.,

18 month'eriods).

The licensee

, provided the following results (i.e., pints of oil added to the up-*

per/l.ower reservoir):

.. RCP

CYCLE 13

3A '2/0

.3B.,

0/0

3C

0/0

18

M2.4

This data

was retrieved

from

PWO search.

Containment

inspections

in the

vicinity of the

RCPs have not noted

any oil buildups.

Although the data

was not researched

for Unit 4, system engineering

and maintenance

personnel

stated that the oil usage

was similar.

The inspector also reviewed the cur rent status of the 48 RCP.

The 48

RCP has

had an historical oil level alarm which requires

operator

response

and enhanced

monitoring.

The licensee

determined that this oil

level alarm was due to a design deficiency.

Further.

the 48

RCP has

a

slight seal

housing flange leak which is being monitored by a remote

camera

set

up in the Cable Spreading

Room.

CR No.97-790 addressed

this

issue of the noted dry boric acid.

Operators

monitored this seal

housing every four hours.

No active leakage

has

been noted.

The

licensee

intends to make repairs

during the September

1997 scheduled

refueling outage.

Conclusions

The inspector

concluded that the licensee

adequately

monitors

RCP oil

consumption.

and that issues

associated

with the 48

RCP were appropri-

ately documented

and dispositioned.

Hi h Head Safet

In 'ection

HHSI

Pum

Status

62707

Ouring the period, operators

noted minor pump casing leaks

on the 4A and

38 HHSI pumps.

The licensee quantified these

leaks by running each

pump

on full flow recirculation.

The leakage

was calculated to be 330 cubic

centimeters

per hour (cc/hr).

The licensee verified that this was

acceptable

per pump vendor requirements,

and the leakage

was less than

the total

as stated in the

UFSAR Table 6.2-12.

Licensee corrective

actions

included minimizing pump use for cold leg accumulator fills,

issuing

a problem status

summary,

placing an information tag on the

control

room switches,

performing daily leakage monitoring,

and

developing

an action plan to perform repairs.

On June 5 1997, the licensee identified that the 4A HHSI had two

failures in the past

18 months.

These were maintenance

preventable

functional failures

and therefore placed the component in category a(1)

of 10 CFR 50.65,

Maintenance

Rule.

The two,fai lures were previous

casing

leaks (reference

CR No.97-613 and

NRC Inspection

Report

Nos.

50-

250.251/97-03

and 04),

and breaker failure (reference

CR No.97-846).

The enhanced

monitoring per the Maintenance

Rule was documented in a

separate

CR No.97-955.

.The i.nspector. reviewed the above mentioned

CRs,

observed. HHSI pump

"'. testing in the field, veri.fied .corrective actions,

and discussed. these

.issues with maintenance,

engineering,

and operations

personnel.

The

inspector intends to review these

issues

in future inspections.

19

H2.5

Process

Radiation Honitorin

PRH

S stem

62707

During the period, the inspector

noted recent

PRH system problems

including the following:

Three failures

(RD-18, RD-3-15.

RD-3-17A) associated

with the

amphenol

connectors that connect the field wiring (from the

detector) to the control

room drawer.

Steam jet air ejector

(SJAE) particulate,

iodine, noble gas

(SPING) monitor fai lures due to high moisture content,

material

compatibility, and environmental

conditions.

Each failure was addressed

by an associated

PWO and

CR.

Collective

assessments

and Maintenance

Rule applicability determinations

are

pending.

In addition, the licensee is addressing

longer term corrective

actions.

The inspector

reviewed this issue with operations

and engineering

personnel.

TS and Offsite Dose Calculation

Manual

(ODCH) requi red

action statements,

and alternative

sampling requirements

were verified.

The inspector expressed

concerns

regarding

PRH system reliability and

intends to review this in a future inspection.

H2.6

Plant Material Condition and Housekee in

Issues

62707

During the period, the inspectors

noted

a number of equipment issues,

some degraded

plant material conditions,

and examples of poor housekeep-

ing including the following:

Poor water leak tightness of rooms

and equipment

(section 01.3),

AFW Unit 4 train two flow control valves

and piping in need of

preservation

(section H1.2),

Unit 4

PORVs leakage

(section

M2. 1),

Units 3 and 4 pipe and valve rooms poor general

area conditions

and housekeeping,

Process

radiation monitoring system failures and related

issues

(section H2.5),

.... Untimely'cleanup after. a work performance in the auxiliary build- .'-

. ing (section 01.2),

and

4A and

3B HHSI pumps casing

leaks (section H2.4).

These

issues

were discussed

with the individuals at the time of

discovery by the inspectors,

and with plant management.

The inspectors

,concluded that additional licensee attention in this area

was warranted.

20

H8.1

E2

E2.1

Miscellaneous

Maintenance

Issues

Closed

VIO 50-250 251/96-06-02

and

LER

50-250 251/96-08

90712 and

92902

A related

LER (50-250,251/96-08)

was also submitted concerning this pre-

conditioning of the diesel

fuel priming system.

The violation response

(L-96-185) and

LER corrective actions

were reviewed

and verified to be

appropriate.

The

LER and VIO were closed.

En ineerin

Engineering

Support of Facilities and Equipment

Fire in 4A Motor Gener ator

MG

Set

Ins ection

Sco

e

37550

On March 4

~ 1997, control

room operators

received fire alarms for the

inverter

rooms

and the cable spreading

room.

The fire was in the 4A

control rod drive motor-generator

(HG) set.

The inspectors

reviewed

Condition Report

No. 97-0286

and attended

the Plant Nuclear Safety

Committee

on this subject.

The inspectors

evaluated

engineering's

support for the other plant organizations for determining the root

causes

and

recommending corrective actions.

The event

was reviewed in

NRC Inspection

Report

No. 50-250,251/97-03.

Observations

and Findin s

The

CR documented

the evaluation of the following three aspects

of the

subject event:

(1) the fire and fire response;

(2) effects of the dry

chemical

from the fire extinguisher

on the

HG set

and surrounding

equipment;

and,

(3) the root cause for the bearing failure on the

HG

set.

During this review, the inspectors

concentrated

on the third

aspect

(the root cause of the bearing failure).

The bearing failed prematurely after

11 months of service.

These

bearings

normally run 36 months without,failure before replacement.

Several possibilities

were considered for the root cause of the bearing

fai lure.

One was the possibility of an increase

in vibration for the

Unit 4

HG set

as

a result of stopping the Unit 3

HG sets.

Vibrational

measurements

eliminated this as

a potential root cause.

A second potential root .caUs'e,

an insufficient lubrication problem,

was

considered. 'ecords

show'hat the beating

had been lubricated in the

same

manner. as were all- of. the

MG set bearings

si.nce

1983.

The licensee

used less lubri'cation than the vendor

recommended with apparent

success

since

no fai lure ha'd occurred'in this. type of bearing in the last

10

years.

Approximately 1.5"cubic inches of grease

was found in the

inboard bearing cavity even after the fire.

The licensee'ha's'ecided.to.

increase. the amount of grease .injected in the bearing during preventive

21

maintenance activities.

The licensee realized

and industry experience

showed that the bearing

can fail from too much grease,

as well as too

little.

A third potential root cause

was

a materials or fabrication problem.

Vibration analysis

taken within a month of the bearing failure did not

indicate any degradation.

The inspectors

reviewed the metallurgical

analysis

and discussed

the results during

a tour of the

FPL

Metallurgical Laboratory facilities.

A review of the information in the analysis

showed the material

used

was

alloy 52100 which was

a correct material selection.

The thickness

and

color of the oxide on the fracture face of a crack across

the inner ring

of the bearing indicated that the temperature of the part had exceeded

650 degrees

Fahrenheit

(F).

Normally these

bearings

are not operated

above

300 degrees

F and lubrication usually breaks

down between

400 to

450 degrees

F.

This information indicated that the final fai lure of the

bearing

was caused

by lack or fai lure of lubrication.

In addition, the

extreme friction between the unlubricated metal parts generated

enough

heat to temper

(soften) the normally hardened

steel structure.

No

abnormal fracture face characteristics

nor abnormal structures

or

chemistry were noted.

c.

Conclusions

'he

exact root cause of the failure could not be determined.

However,

the inspector

independently

reviewed the professional

approach

for

evaluating bearing fai lures in accordance

with Volume ll American

Society of Materials

(ASM) Handbook,

Failure Analysis and Prevention,

Section,

Failure of Element-Roller

Bearings)

and concluded that the

licensee

had performed

a thorough

and proper root cause analysis.

Also,

the inspectors identified a positive finding that the licensee

had

a

good Metallurgical L'aboratory with excellent equipment

and that Turkey

Point was proactive in requesting

the laboratory's

services

for failed

components

or parts.

E2.2

Retubin

the 3A Com onent Coolin

Water

Heat

Exchan er

Ins ection Sco

e

37550

The licensee

was retubing the

CCW heat exchangers

on Unit 3 and has

completed the 3B heat exchanger

(reference

NRC Inspection

Report

No. 50-

250,251/9?-01).

The -i'nspectors

observed part of'he work that was

occurring during this inspection period on retubing the.3A heat

exchanger,

reviewed the engineering

support;"and.reviewed

the

..

engineering

decisions'.supporting.

the project.

.

4

b: 'bservations"and,Findin

s

r

On May 10,

1997, the licensee

removed the Unit 3 3A CCW heat

exchanger

to replace the tubes.

The Turkey Point units each

have three'0 percent.

CCW heat exchangers.

TS'.7.2.b'only requires

two heat exchangers,

thus,

E2.3

the third heat exchanger is an installed spare.

Since no TS action

statement

exists for one heat exchanger to be out of service,

the

licensee conservatively

used

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

The

retubing activity was longer than the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months administrative limit,

therefore the Plant Nuclear Safety Committee

(PNSC) reviewed

and

approved

an extension.

The inspectors

reviewed Condition Report

No. 97-0070,

which was

origina11y written to retube the 38

CCW heat exchanger.

An operability

determination

was

made for all of the Unit 3

CCW heat exchangers

because

of the presence

of foreign material

found in 3B.

The majority of the

material

was from a stainless

steel wire mesh strainer

placed in the

line before

1972 (pre star tup).

In the 3A heat

exchanger

the foreign

material

appeared to be similar.

No unusual

wear or failure of the

tubing could be attributed to the foreign material

and thus

no

operability problems existed.

The inspectors

had no problems with the

logic used for the operability evaluation.

The inspectors

observed

some of the detubing

and retubing operations,

cutting an opening in the shell of the heat exchanger

vessel

to remove

the foreign material,

and

some of the welding to replace the opening in

the shell.

Discussions

were held with the systems

engineering at

various times during the project.

Conclusions

The inspectors

observed that the

r etubing operation

was properly

performed.

Also, the inspectors

observed

strong engineering

support for

maintenance

and operations

during this project with around the clock

shift coverage

by engineering

personnel.

This was another

example of the excellent support provided by

engineering for the operations

and maintenance

organizations.

The

inspector

noted that engineering

support

was routinely requested

by

these organizations

on

a daily basis for work planned or in progress.

Reactor

Coolant

Pum

RCP

Motor Oil Collection

S stems

37551

On June

18,

1997, at 3:05 p.m.. the licensee

made

a one hour

ENS call to

the

NRC per

10 CFR 50.72.

The licensee

concluded that

RCP oil

collection system for both units

may not meet

NRC fire protection

requi rements

per

10 CFR 50, Appendix R, Section III.O.

During

, 'reparation for a modification to the lube oH fill connection,

'dditional potential

leakage. sites were identi.fied that were not

collected

by the oil collection..system:

'Because

these potential'leakage...

.sits were either unpress'urized

or low pressure,

any postulated

leakage

'ould

not be in 'quantities sufficient to sustain

a fire.

Therefore,

the

~ .l.icensee

concluded that probabi lity and'ffect of a postulated fire were

.

minimal.

The oil collection system

has

been evaluated

and determined to

be operable,

based

on the extremely small likelihood of a fire. " No

.actual

leakage

has

been observed

from any of the uncol.lected potential

E4

E4.1

E5

E5.1

23

leakage sites.

PNSC reviewed

and approved

CR No.97-098 which was

written to document the condition.

The inspector

reviewed the

CR,

UFSAR section 9.6A, the above

requirements'nd

other problems to HP supervision.

These

problems were reported through

the

CR process.

Release of Radioactive Contaminated

Paintin

E ui ment To Unrestricted

Areas

Ins ection Sco

e

83750

This area

was revi ewed to evaluate the ci rcumstances

concerning the

release of contaminated

equipment

from the Turkey Point site.

Observations

and Findin s

Background

The licensee utilized special painting equipment to paint the surfaces

of the reactor cavity walls during refueling outages.

The paint sealed

the contamination

on the walls to minimize the spread of radioactive

materials in the work ar ea.

The paint also helped decontaminate

the

walls when it was later stripped from the walls.

The painting equipment

was

owned by Power Systems

Energy Services,

Inc (PSESI).

The vendor

needed the equipment for

a simi lar job at the Braidwood nuclear station

in Illinois.

On March 25,

1997. licensee

personnel

logged the release of the painting

equipment

on licensee

form "RCA Release

Log."

The equipment listed

included two PSESI paint pumps with hoses

and fittings.

The equipment

was shipped to the vendors'acilities

in Altamonte

Springs.

Florida.

From there the equipment

was taken to a paint supply

company for maintenance.

The maintena'nce

was performed

by Lee Patterson

Company in Orlando, Florida.

The equipment

was shipped

from the

maintenance

shop to the Braidwood nuclear station.

Neither

PSESI or Lee

Patterson

possessed

a radioactive materials

license.

Braidwood personnel

surveyed the equipment prior to it's use at thei r

facility on April 30,

1997,

and found fixed contamination

up to 5,000

dpm and .smearable

contamination

up to 3,000 dpm/100 cm'.

The licensee

was notified of the contamination

problem on May 1,

1997.

,. The licensee. notified the State of Florida Department of Health,.Bureau

of Radiation Control

(BRC), the

NRC resident inspectors.

and made

a

report in accordance

with 10 CFR 50.72(b)(2)(vi) on May 1,

1997.

Representatives

from the Florida

BRC surveyed the

PSESI

and

Lee Paterson ..

facilities where the equipment

had been stored

and worked on.

The

BRC

also surveyed the vehicle used to carry the equipment to maintenance

30

facility.

The state did not find any radioactive contamination

except

for a file box marked "paint sprayer" which had radiation levels twice

background.

The state confiscated the box.

The maintenance facility

transferred the paint hoses

and spray guns to Braidwood in a paste

board

box.

The Braidwood staff surveyed the received

equipment

and measured

the

following radioactivities:

Spray nozzle

5,000 dpm/100 cm'ixed

Spray gun hose 3,000 dpm/100 cm'mearable

and

5,000 dpm/100 cm'ixed

Spray tip

2.000

dpm/100 cm'ixed

'Spray gun

1,000 dpm/100 cm'ixed

Spray gun

1

~ 000 dpm/100 cm'ixed

Equipment

bag <1,000 dpm/100 cm'mearable

Paint sprayer

1,000 dpm/100 cm'ixed on valve

The paint spray gun hose

was the only equipment the Braidwood staff

found having smearable

contamination.

Observations

The licensee initiated

CR 97-0828

on Hay 1,

1997. to cause corrective

actions for the release of'he contaminated

materia1.

The cause listed

on the

CR was "Survey of material using current equipment

and techniques

does not assure

100 percent detection of <5000 dpm."

The

CR also stated

that the Turkey Point release

methods

could have missed the levels of

contamination

reported

by Braidwood.

'Title 10 CFR Part 20.1501(a),

requires,

in part, that each licensee

make

or cause to be made,

surveys that

may be necessary

for the licensee to

comply with the regulations

and are reasonable

under the circumstances

to evaluate the extent of concentrations

or quantities of radioactive

material

and the potential radiological

hazards that could be present.

The regulations

applicable to nuclear

power reactor licensees

do not

rovide for release of materials for unrestricted

use that are

known to

e radioactively contaminated at any levej.

The licensee's

fai lure to

detect 3,000

dpm smearable

and up to 5,000

dpm fixed radioactive

contamination

was identified as

a violation of 10 CFR Part 20. 1501

requirements (first example).

The item is tracked

as

VIO 50-250,251/97-

06-02, Failure to Control Licensed Byproduct Haterial

and Hake Adequate

Contamination

Surveys

(Painting Equipment

Released

from RCA).

The, inspectors

noted the following vulnerabi 1'ities .concerning the

licensee's

RCA exit surveys:

" The licensee did .not routinely count smears for loose contamina-

tion on materials exiting the

RCA with sample counting systems.

. The licensee relied exclusively on the thin window GH detector

for

.both fixed and smearable

contamination.

The licensee

had capabil-

ities to detect

much lower levels of smearable

contamination

(less

31

than

200 dpm/100 cm') with counting instrumentation

but relied on

detection

methods only capable of detecting

1,000 dpm/100 cm'.

The licensee's

procedures

permitted the release of porous materi-

als without any additional precautions.

A primary RCA exit and survey point did not have any counting

instrumentation.

was not enclosed,

or well lighted at night.

During the outage the licensee

assigned

and rotated multiple

vendor

HP technicians

during

a shift to survey materials exiting

the

Radiation survey records

were not required for

HP surveyed

items

exiting the

c.

Conclusion

The inspectors

found that the licensee's

efforts in detecting

and

measuring

contamination levels

on items released

from the

RCA were

practical

and

common.

However, the licensee's

administrative controls

and measurement

techniques

were not good enough to detect the released

contaminated material.

One violation concerning the release of contami-

nated material

was identified.

Rl.3

Release of Radioactive Contaminated

Com anent Coolin

Water Heat

Exchan

er

Tubes

To Unrestricted

Areas

a.

Ins ection Sco

e

83750

This area

was reviewed to evaluate the circumstances

concerning the

release of contaminated

Component Cooling Water

(CCW) Heat Exchanger

(Hx) Tubes

from the Turkey Point site.

b.

Observations

and Findin s

BACKGROUND

The

CCW system is the heat sink for many plant components

including the

Residual

Heat

Removal

Loop, Chemical

Volume Control System,

Spent

Fuel

Cooling Loop and various Reactor Coolant System components.

Makeup

water for the

CCW system is supplied from the water treatment plant

through .the sUrge tank.

The

CCW cools systems

from the shell side of

, various

Hxs.

The

CCW system "is cooled by 'ntake Coolant Water System

, entering the tubes. of, the

CCW Hxs.

'n

discussions with licensee

personnel

the inspector

learned the

. "

licensee

was replacing tubes in.all. 3 of the Unit 3

CCW Hxs in 1997.

The*CCW tubes were. 18 gauge aluminum/brass

tubes

having

a saltwater

service life of approximately .15.years

and

had .been in service'or

nearly

25. years.

Some of .the Unit 3

CCW tubes

had been replaced,.in

1991,

leaving about 1,425 tubes to be replaced in each of the Unit 3

I'

32

Hxs.

The licensee

replaced the tubes

on 3B

CCW Hx in February

and was

replacing the 3A CCW Hx in May and

3C

CCW Hx in June,

1997.

The licensee

surveyed the 3B

CCW Hx tubes for contamination

using thin

window Geiger Muller

(GM) detectors

and Micro-R Meters.

The licensee

also checked for loose contamination

using masslin cloths

and thin

window GM detectors.

The licensee

surveyed

100 percent of the external

tube surfaces

for approximately 2/3 of the

3B Hx tubes.

Since the

licensee

had not identified any contamination

on the tubes during those

surveys,

the licensee

relaxed the survey methods for the last third of

the 3B

CCW Hx tubes.

The licensee

surveyed portions of the remaining

tubes

released.

The 100 percent

survey process

had been time consuming

taking two Health Physics Technician

(HPTs) approximately

2 weeks with

some overtime to complete the task.

SEQUENCE

OF

EVENTS

The inspectors

learned that the licensee

may have released

some contami-

nated

CCW Hx tubes for unrestricted

use.

Through interviews with

licensee

personnel,

review of records,

observations.

and radiation

and

contamination

surveys the inspectors

determined the following.

On Friday May 30,

1997,

a Mechanical

Maintenance

(MM) foreman directed

MM personnel

collect

a sample the 3A CCW Hx tubes.

The mechanics

were

instructed to cut portions of the

CCW tubes to fill two one liter

marinelli beakers.

The samples

were taken

and delivered to the

HP

counting

room that day.

A HPT working in the counting

room counted the

samples

on

a Multi Channel

Analyzer

(MCA) for 1,000 seconds.

No

radioactive nuclides were identified on sample

ID M1971390.

However,

byproduct radionuclide Co" was identified on sample

ID M2971389.

The

sample

was counted

a second time on another detector

and was identified

as sample

ID M1971391.

The second analysis of the sample again identi-

fied the presence

of radionuclide Co".

The

MCA was not setup to

quantify radioactivity on metals in a liter marinelli.

The samples

and

analysis

were made to provide documentation that the

CCW Hx tubes

were

free of all byproduct materials.'owever,

two analysis

reports clearly

identified the presence of Co" activity and listed the "quantities"

as

1.78 E-4 pCi/1

and 1.84 E-4 pCi/l.

-The

HPT that had counted the samples

reported the three

sample analysis

reports

(one negative

and two positive) were clipped together

and placed

on the Health Physics Shift Supervisors

(HPSS) log book on the HPSS's

desk that afternoon.

On the following Monday, June 2,. 1997,

an

HP supervisor-reported

finding ..

only. the 3A CCW Hx sample analysis. report that had not identified. any

'adioactive

material

on. the

CCW tubes.

Based on'the single sample

an'alysis. of the 3A CCW Hx tubes

and the survey history of the

3B

CCW Hx

tubes in February,

1997, licensee

management

reduced the survey

i equire-

ments for the free release of remaining Unit 3

CCW Hx tubes.

.33

The licensee

documented

the decision in inter-office correspondence

to

Quality Assurance

(QA) 1000 File dated

June 3,

1997.

The

memo

authorized the free release of the remaining Unit 3

CCW heat exchanger

tubes

based

on the point of origin and the following radiological

analysis:

2.

3.

4

5.

A two year chemistry history of this system

(no radioactivi-

ty);

Isotopic analysis of random tube samples.

(no activity);

Approximately 50 percent of the estimated

1500

CCW tubes

have

been

smeared

and direct frisked with no radiological

findings;

100 percent masslin wipedown;

and

Micro-R Meter of Tubes.

On Tuesday

June 3,

1997, the licensee

began releasing the 3A CCW Hx

tubes.

The HPTs assigned to survey the tubes

received verbal guidance

to masslin

(smear for loose radioactive contamination)

about

25 percent

of the external

surfaces

and to direct frisk about

25 percent of the

surfaces

with a Micro-R Meter

and

a thin window GM detector.

The HPTs

surveyed portions of the tubes

on the wagon inside the

RCA and

MM

personnel

passed

the tubes

though

a chain link fence

(RCA Boundary) to

another

MM person to load on another

wagon just outside the

Other

HPTs passing

the work location and observing the survey

and

release

process

challenged

the adequacy of the survey process with the

HPTs surveying the tubes.

The HPTs surveying the tubes

were provided

a

copy of the June 3,

1997,

memorandum permitting the release

process.

The licensee

released

approximately

one half of the wagon that day

(approximately

350 tubes).

The tubes

were removed from the protected

area

and

dumped

on the ground

away from the site on the licensee's

Land

Utilization Area.

On Wednesday

morning June 4.

1997,

HPTs were surveying the accessible

surfaces of the remaining tubes

on the wagon inside the

RCA and waiting

'n the

MM support personnel

to arrive.

The

HPT that had counted the 3A

CCW Hx tube samples told the

HPTs that the tubes

should not be released

since they were contaminated with Co".

This information was passed

on

to HP management

and no additional

CCW tubes

were released.

The

licensee

began

an informal investigation.

The licensee

located the

sample having positive Co" and counted the sample again.

The recounts

continued to see the Co" in the sample.

The licensee

reported taking

and analyzing additional

samples

f'rom tubes .still remaining in the

RCA

'nd..that

no Co"'contamination.was

identified in those-samples.

I

"

On. Thu'rsday June '5, 1997,'he

licensee. dispatched

two Mechanical'

Maintenance

(MM) pers'onnel

to the area where the tubes were dumped to

pickup the tubes

and retur'n.then to the site.

However,'he tubes were

.

'ot

there.

The

MM personnel

foreman'also verified the tubes

were not at

the site later that day.

The licensee

reported the decision to recover

the tubes

from the Land Utilization Area was

a precautionary

measure,

in

that, the staff did not believe that there

was any contamination

on the

34

tubes.

As

a result. the licensee

decided that the recovery of the

released

tubes

was not necessary.

The licensee did not know when the tubes were removed from the land

utilization area during the period of June 3-5,

1997.

The licensee did

not receive receipts for items picked up in the area

The person that

icked up the tubes for the salvage

company reported that

he periodical-

y visited the area

where the tubes

had been

dumped for pickup of scrap

metal

and he could not remember the date

he picked up the" tubes.

On Friday.

June

6.

1997. the licensee

was satisfied that the 3A CCW Hx

tubes

were free of radioactive contamination.

The licensee

planned to

continue with the release of the tubes.

However, the licensee

reported

that no additional

tubes

were released

from the site.

On Wednesday

June

11 '997,

a conference call with plant staff was

made

to discuss

the release of CCW Hx tubes with NRC inspectors.

During the

call the inspectors

learned that the licensee

believed that they had not

released

any contaminated

material

and that they had found

a sample

contaminated with low level Co" but that they believed the sample

had

been contaminated.

The inspectors

also learned that the licensee

had

not initiated

a Condition report

(CR) and had not notified the State of

Florida oi'he incident.

The inspectors

informed the licensee that the

NRC may notify the State of Florida of the issue.

The Region II Staff

made

a courtesy call to the State of Florida the same afternoon

and

reported that

some very low level contamination

may have been released

and taken to a salvage yard in Dade County, Florida.

The State reported

that they would probably visit the scrap yard for radiation surveys

and

sample the tubes for the presence of low level radioactive

contamination.

The inspectors later determined that the licensee

initiated

a

CR 97-0985 following the conference call.

On Thursday,

June

12,

1997, the State of Florida Department of'ealth,

Bureau of Radiation Control

(BRC). visited the property of salvage

company

and

made radiation surveys of'he

CCW tubes there.

The State

personnel

did not detect

any radiation above background during surveys

of the materials.

The State also obtained

a sample of the

CCW tubes for

analysis in thei r laboratory.

On Friday, June

13,

1997, the State of Florida

BRC reported

a low but

measurable

quantity of Co" had been identified in the

CCW tube sample

they had taken at the scrap yard.

The level of radioactivity detected

was approximately 2.73 E-2 pCi/g:.

On Monday'June

16,

1997, the licensee

sent personnel

to the salvage yard

to retrieve the

3A. CCW tubes..

When. licensee

personnel

ar'rived the tubes

we'r'e already 'in a dumpster

and on

a fork lift.

The dumpster .was empti,ed..

into the licensee's

truck.

According. to licensee

personnel

the salvage

~ yard was small

and there did. not appear

to be any. other material'in the

yard resembling the

CCW Hx tubes.

The tubes

were returned to the site

'

and placed

i,n

a locked fenced .area

adjacent to the

0

FINDINGS:

35

On Thursday June

26,

1997,

NRC inspectors

surveyed portions of the

returned

CCW tubes with thin window GM detector

and

a Micro R meter and

did not identify any contamination levels greater than background.

The

inspectors

also surveyed

a few of the tubes with 100 cm'mears.

The

smears

were counted

on

a low background counting system

and one the

smears

had contamination of 152 dpm/100 cm'.

The rest of the smears

(9)

did not detect

any contamination in excess of 26 dpm which was the

Minimum Detectable Activity (MDA) for the counter utilized.

The

inspectors

also sampled the tubes

and had them counted

on

a licensee

MCA

for 1,000 seconds.

The

MCA identified Co" on the

NRC inspector's

sample.

-The inspectors

concluded the tubes

were contaminated with very

low levels of measurable

byproduct contamination.

Title 10 CFR Part 20. 1801 required the licensee to secure

from

unauthorized

removal or access

licensed materials that are stored in

controlled or unrestricted

areas.

Title 10 CFR Part 20. 1501(a),

required,

in part, that each licensee

make

or cause to be made,

surveys that

may be necessary

for the licensee to

comply with the regulations

and are reasonable

under the circumstances

to evaluate the extent of concentrations

or quantities of radioactive

material

and the potential radiological

hazards that could be present.

The regulations applicable to nuclear

power reactor licensees

did not

rovide for release of materials for unrestricted

use that are

known to

e radioactively contaminated

at any level.

The licensee's

failure to

control licensed materials

and

make adequate

radiation surveys

was

identified as

a violation of 10 CFR Part 20.1801

and 20.1501 require-

ments

(second

example).

The item is tracked

as

VIO 50-250,251/97-06-02,

Failure To Control Licensed Byproduct Materials

and Make Adequate

Contamination

Surveys

(3A Component Cooling Water

Heat Exchanger

Tubes

Released

From The RCA).

The inspectors

reviewed the licensee's

management

controls for the

release of the 3A CCW tubes.

The inspectors

determined that there

was

a

breakdown in management

controls

and communication associated

with the

release of the contaminated

3A CCW tubes.

The

HP supervision staff

reported that they were unaware that two samples

had been requested

by

MM supervision

and that they had not received the two sample analysis

identifying the presence. of Co".

HP personnel

had not directly

'supervised

the sampling'f the

CCW tubes.

It appeared that the nuclide

activity. reports

having identified the'resence

of low level

contamination

on

CCW tubes

had been lost.

'One of the bases

for. the 'reduced

survey req'ui'rements

used. in releasing...

. the

CCW tubes .was that there

had not been

any radioactivity detected in

the Unit '3

CCW Hx tubes within the last two years.

However, the"

. inspectors

found that was technica] ly. incorrect in that contamination

had been identified in the

CCW system in August 1995:

.The licensee

. "-

routinely sampled the system monthly and when radioactivity was found in

36

the system, it was sampled weekly as long as radioactivity was

identified.

The licensee

had last seen radioactivity in the

CCW system

in August 1995.

Concentrations

of Na" 1.38 E-7 yCi/cc and Cs'" 2.24 E-7

pCi/cc were identified that month.

The inspectors

reviewed licensee's

responses

and corrective actions

concerning,

the release of the

CCW tubes.

The inspectors

found

management's

response

and assessment

concerning the release of the

contaminated

3A CCW tubes

was slow to recover the contaminated

tubes

and

it did not address

management

control failures or the missing sample

analysis reports.

The inspector determined the following:

Management failed to initiate

a formal review of the problem unti 1

prompted

by NRC review of the event;

Management failed to identify the breakdown in management

controls

concerning the sampling

and review of the sampling results in

thei r formal review of the event;

Licensee

had not identified missing survey records in the

corrective action program;

When there

was

some question concerning the presence of

radioactive byproduct material

on the tubes,

licensee

management

was slow in making the decision to retrieve the tubes;

and

When the licensee

management

believed there were reasons

to

retrieve the tubes

and found that they had been

removed from the

site the licensee

made

no further attempts to retrieve the

materials.

The licensee's

CR 97-0985 identified the problem as "Uncontrolled

composite sampling of 3"A" CCW Hx tubes,

and timely notification of

survey results; for determination of release of material

fro'm the

Resulted in an allegation that material

was improperly released

from the

RCA"

The

CR identified the cause

as

"Use of RCA identifiable tools to

cut up tubes for sample analysis,

and improper notification of survey

results."

The licensee

suggested

the possibility that

a contaminated

tube cutting

tool had contaminated

the tube samples with Co" since the tool was

identified as

a tool used for work in the

However, the inspector's

samples

of. the tubes also identified Co" and smearable

beta

and

gamma

'contamination. 'he

CR also faulted the timeliness of the technician "s

repor't to. management. that the tube. s'ample contained -Co".

The inspector

."

determined that. the counting

room HPT had followed, routine'piactices

in

. taking the sample analysis

reports to HPSS office. several

days prior to

.the. licensee.'s

decision to release

the

CCW tubes.

During the inspection licensee

management

reported that

MM personnel

were dispatched

on Wednesday

June 4,

1997, to retrieve the tubes.-

However, the inspector determined'through'interviews

with licensee

37

personnel

that persons

were not dispatched to the Land Utilization Area

to retrieve the tubes until Thursday June 5,

1997.

The licensee

reported that all of the

CCW tubes

had been returned to the

site.

However, several

persons

interviewed believed that the quantity

of tubes

returned to the site were less than the quantity released

from

the site.

The inspectors

were unable to determine whether the quantity

returned

was equal to the quantity released.

According to licensee

ersonnel

when the tubes

were released

they were approximately

15 foot

ong.

None of the returned pieces

observed

by the inspector were

15

foot in length.

Most of the tubes

were approximately

30 to 50 inches in

length and several

had been bent.

The licensee did not weigh the

material in or out.

The inspectors

were unable to determine whether all

of the tubes

had been returned to the site.

, The salvage operator

reported that all of the tubes

picked up at the dump site had been

returned to licensee

personnel.

The licensee

did not plan to unconditionally release

the remaining

tubes.

The licensee

planned to ship the tubes to a vendor for

processing

and disposal.

Conclusion

P1

P1.1

a

b

There was

a breakdown in management

controls

and communication

associated

with the release of the contaminated

3A CCW tubes.

Management's

response

was slow to retrieve the contaminated

tubes

and

their assessment

concerning the release of the contaminated

3A CCW tubes

did not address

management

control failures.

A violation was identified for fai lure to control licensed

byproduct

materials

and

make adequate

contamination

surveys of 3A CCW tubes

released

from the

Conduct of EP Activities

Hurricane

Pre ar ati ons

Ins ection Sco

e

71750

The inspectors

reviewed

and discussed

with the licensee the program and

procedures

associated

with hurricane preparedness.

Hurricane season

spans

the months of June through November with the most intense activity

expected to occur between

August and October

.

~

C

I

Observations 'and. Findin s

There are procedures,'Ms.

and other preparatory

processes

that the

licensee. performs at the onset of each hurricane

season=:

Additionally,

there are procedures

that the licensee

would implement

upon declaration

of a hurricane watch or warning.

The licensee

has the .following.

procedures

in,place'to ensure

adequate

preparation

due. to a hurricane:

38

Procedure

O-ONOP-103.3,

Severe

Weather

Preparations.

provides

instructions for the preparation of the site for severe

weather

conditions not resulting in implementation of the Emergency

Plan.

This procedure

would be entered

upon the notification of a

tropical Storm Warning or

a Hurricane Watch which includes the

Turkey Point site.

(A Hurricane Watch is declared if a hurricane

is located

between

24 to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from and is approaching

the

United States

coast.

A Hurricane Watch area

includes

approximately

100 miles on either side of the expected landfall

location.)

Instructions

and guidelines for preparing, controlling,

and

recovering the plant following activation of the Emergency

Plan

for a natural

emergency

are provided in procedure

EPIP-20106,

Natural

Emergencies.

This comprehensive

procedure

addresses

tornadoes

and hurricanes,

but is to be used for any severe

weather

disturbance

which results in the activation of the Emergency

Plan.

It also contains specific guidance for coping with the possible

flood conditions associated

with more intense hurricanes.

This

procedure

would be entered in advance of a Hurricane Warning.

A

Hurricane Warning is declared if a hurricane is located

between

12

and

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

from and is approaching

the United States

coast.

A

Hurricane Warning area

includes approximately

50 miles

on either

side of the expected landfall location.

Procedure

0-SMM-102. 1, Flood Protection Stoplog

and Penetration

Seal

inspection,

is uti lized by the licensee to verify operability

and adequate

inventory of flood protection equipment.

Security force instruction SFI-3002,

Hurricane Preparedness,

provides guidance for security activities in preparation for,

during,

and following hurricane threats or actual conditions.

The

FPL Nuclear

Power Plant Recovery Plan is an

FPL corporate

document which establishes

a pre-planned

organization

and action

plan to recover from a nuclear

power plant emergency

and minimize

unfavorable

impact on the

FPL plants

and the public.

Procedure

EP-AD-009, Hurricane Season

Preparation,

is an

administrative directive which is implemented prior to each

hurricane

season

(e.g.,

June

1)

In additi.on, the licensee

has prepared

a detailed hurricane schedule

flow..chart (P-2) using thei r. corpor.ate

schedule

programming capability.

This schedule

sequences,

'documents,

and tracks all necessary

steps to be

completed prior to, during,

and after

a hurricane strike;

Licensee actions relative to the units are:

Cate or

1 or 2 Hurricane

39

Proceed to Mode 3 (Hot Standby)

per the requirements of'80

per

NUMARC 87-00 (reference

L-90-338 September

21,

1990).

Cate or

3

4

or 5 Hurricane

Proceed to Mode 4 (Hot Shutdown)

and maintain

RCS

Tavg

between

350-343'F to assure

AFW operating

steam pressure

(785 psig).

The licensee's

preliminary preparations

for hurricane

season

have been

completed.

The satellite up-link communication capability is on-site

and ready for use,

and the stoplog walkdown inspections

have

been

performed.

The licensee

has also procured

and stored non-perishable

food supplies

and the storm supply inventory for preparatory

actions

required

by procedures.

Prior to the onset of a hurricane.

these

items

would be moved to the designated

storage

areas.

The inspector

reviewed

the licensee's

procedures,

storm stock inventory lists,

and

PWOs

regarding the flood protection stoplog inspection

and various floor

drain inspections.

In addition, the licensee

conducted

a Table Top

drill to test their hurricane schedule

implementation.

Conclusions

The inspector concluded that the licensee

has

been proactive in the area

of hurricane preparedness.

Conduct of Security and Safeguards Activities

Access Authorization

Ins ection Sco

e

TI2515/127

The inspector

reviewed

a portion of the licensee's

Access Authorization

Program

(AAP) to determine if the requirements of 10 CFR 73.56 were

being met,

as committed to in the

NRC approved Physical Security Plan.

Observations

and Findin s

The inspector

randomly selected

ten

AAP records to review and determine

if individuals'enied

unescorted

access

to Turkey Point were properly

notified and otfered

an appeal to the decision.

The inspector noted the following circumstances

during the file review:

On February

17,

1997',

an individual began the process

of

r einstating

'hei

r access

authorization

from St. Lucie to Turkey Point.

However,

due

to previous fitness for'uty 'concerns,

the individual would be entered

.into the

FP8L Conditional Access

Program..

Upon completion of the..

individual's.pre-access

chemical

screening,

which was negative,

and

additional

AAP requirements,

the individual was informed he would be

denied unescorted. access

on February,28,

1997... .."

40

The inspector interviewed licensee

representatives

and determined that

the individual would be denied

access

for the current outage at Turkey

Point.

The Site Vice President

made

a conservative

determination that

due to the short time span of the outage,

conditional

access

for

incoming employees

would be eliminated,

unless

no one else could perform

the job.

However,

upon further documentation

review, the inspector

identified that the individual was offered to withdraw his unescorted

access

authorization

request.

The withdrawal

was completed

on Harch 7,

1997, confirming that

FP8L did not deny his request for access

and the

individual would be eligible for future employment at

FP8L facilities.

c.

Conclusions

The inspector

determined that the licensee's

AAP with respect to denial

of unescorted

access

and the appeal

process

met the requirements

of

10 CFR 73.56.

S6

Security Organization

and Administration

S6. 1

Securit

Mana ement

Chan

es

71750

During the period, the Security Supervisor left Turkey Point to become

the Security

Manager

at Florida Power Corporation's

Crystal River 3

Nuclear facility.

Hr. John

Kirkpatrick was appointed

as the interim

replacement

supervisor.

Fire Protection Staff Training and gualification

F5.1

Fire Dr i 1 1 s

71750

The inspectors

observed fire drills conducted

by the site fire protec-

tion organization

on June

2 and 11,

1997.

A simulated fire in the

vicinity of the Unit 3 turbine lube oil tank was responded to by the

five member fire brigade.

In addition, security.

chemistry first aid,

site medical,

and operations

personnel

also responded

as required.

The licensee appropriately provided

a fire drill scenario,

conducted

and

critiqued the drill, and provided immediate

feedback to the partici-

ants.

The licensee

demonstrated

excellent drill conduct

and fire

rigade readiness

for response.

F8

Hiscellaneous

Fire Protection Issues

F8.1

Closed

VIO 50-.250 251/96-11-03

92904

The. issue concerned

a fai.lure. to control, plant design

as required to

meet

10

CFR- 50 Appendix

R cable 'separation

requi rements.

.The licensee

responded

in a letter '(L-96-285) dated November'2,

1996.

The two fire

areas

(Zones

64 and 143) were both addressed.

Corrective actions

included compensatory

measures

and repairs/exemptions

to achieve full

Appendix

R compliance.

These activities are"ongoing and.are part-of the

overall Turkey Point 'thermolag

upgrade project.

.Recent.

meetings'

.

0

41

discussions,

and tours by NRR and Regional

personnel

have been

conducted.

Full compliance

and thermolag

upgrades will be the subject

of future

NRC inspections.

Based

on the completion of licensee actions,

on future licensee activities relative to thermolag,

and on ongoing

NRC

involvement, the violation was closed.

Hang ement Heetin

s

Exit Meetin

Summar

The inspectors

presented

the inspection results to members of licensee

management

at the conclusion of the inspection

on July 2,

1997.

The

licensee

acknowledged

the findings presented.

The inspectors

asked the licensee

whether any materials

examined during

the inspection

should

be considered proprietary.

No proprietary

information was identified.

Partial List of Persons

Contacted

Licensee

T.

V. Abbatiello, Site Quality Manager

R. J. Acosta, Director. Nuclear Assurance

J.

C. Balaguero,

Plant Operations

Support Supervisor

P.

M. Banaszak,

Electrical/l&C Engineering Supervisor

R.

Brown. Health Physics Supervisor

T. J. Carter,

Maintenance

Support Supervisor

B.

C.

Dunn, Mechanical

Systems

Supervisor

R. J. Earl.

QC Supervisor

S.

H. Franzone.

I8C Maintenance Supervisor

J.

R. Hartzog,

Business

Systems

Manager

G.

E. Hollinger, Licensing

Manager

R. J.

Hovey. Site Vice-President

M.

P.

Huba,

Nuclear Materials

Manager

D.

E. Jernigan,

Plant General

Manager

T. 0. 'Jones.

Operations

Supervisor

H.

D. Jurmain. Electrical Maintenance Supervisor

V. A. Kaminskas,

Services

Manager

A.

N. Katz, Mechanical

Maintenance Supervisor

J.

E.

Kirkpatrick, Fire Protection,

EP, Safety Supervisor

J.

E. Knorr, Regulatory Compliance Analyst

G.

D. Kuhn, Procurement

Engineering Supervisor

R. J.

Kundalkar, Vice President,

Engineering and,Licensing

H. L. Lacal, Training Manager

J.

D.-.Lindsay, Health Physics/Chemistry'echnical

Super visor .

.

E. Lyons, Engineering 'Administrative Supervisor

C.

L. Howrey, Licensing Specialist.

H.

N. Paduano,

Manager,

Licensing and Special

Projects

M. 0. Pearce,

Maintenance

Manager

K.

W. Petersen,

Site Superintendent

T.

F... Plunkett,

President,

Nuclear Division

42

K. L. Remington,

System

Performance

Supervisor

R.

E.

Rose,

Work Control Hanager

C.

V. Rossi,

QA and Assessments

Supervisor

W. Skelley.

Plant Engineering

Hanager

R.

N. Steinke,

Chemistry Supervisor

E. A. Thompson'ngineering

Manager

D. J.

Tomaszewski,

Systems

Engineering

Manager

G. A. Warriner, Quality Surveillance

Supervisor

R.

G. West, Operations

Manager

Other licensee

employees

contacted

included construction craftsmen,

engineers,

technicians,

operators.

mechanics,

and electricians.

43

Partial List of Opened,

Closed,

and Discussed

Items

0 ened

50-250.251/97-06-01

50-250,251/97-06-02

50-250,251/97-06-03

50-250,251/97-06-04

Closed

50-250,251/96-13-02

50-250,251/97-06-01

50-250 '51/96-06-02

50-250,251/96-02-02

LER 50-250, 251/96-11 LER 50-250,251/96-08 LER 50-250,251/96-04 LER 50-250,251/96-05

50-250,251/96-11-03

Failure to Follow 18C Surveillance

(section Ml.5)

Failure to Control Licensed Byproduct

Material

and

Hake Adequate Contamination

Surveys

(2 examples)

(section

R1.2 and

R1.3)

URI

RCP Oil Collection System (section E2.3)

IFI

Followup on

QA Audit for Corrective

Actions, (section

E7. 1)

Failure to Follow Radwaste

OP (section

08.1)

Failure to Follow 18C Surveillance

(section Ml.5)

Failure to Adequately Test the

EDGs

(section H8.1)

IFI

AFW Systems

Issues

(section

E8. 1)

LER

Potential

For

PAHH System

Overpressurization

(section

E8.2)

LER

Failure to Adequately Test the

EDGs

(section

M8.1)

LER

Surveillance Testing Reviews per

GL 96-01

(section

E8.3)

LER

Potential

Cross-Tie of Cold Leg

Accumulators

(section

E8.4)

Appendix

R Cable Separation

Requirement

(section F8.1)

44

List of Inspection Procedures

Used

IP 37550:

IP 37551:

IP 40500:

IP 61726:

IP 62707:

IP 71707:

IP 71750:

IP 83750:

IP 90712:

IP 90713:

IP 92700:

IP 92901:

IP 92902:

IP 92903:

IP 92904:

Engineering

Onsite Engineering

Effectiveness of'icensee

Controls in Identifying,

Resolving,

and Prevent

Problems

Surveillance

Observations

Maintenance

Observations

Plant Operation

Plant Support Activities

Occupational

Radiation

Exposure

Inoffice Review of Written Reports

Review of Peri odi c Reports

Onsite Followup of Written Reports of Nonroutine Events at

Power Reactor Facilities

Followup - Operations

Followup - Engineering

Followup - Maintenance

Followup - Plant Support

IP 93702:

Prompt Onsite

Response to Events at Operating

Power Reactors

TI 2515/127

Access Authorization

List of Acronyms and Abbreviations

AAP

AFW

a.m.

ANSI

ARP...

ASM

BRC,'A

CA

cc/hr

CFR..

Access Authorization Program

Auxiliary Feedwater

Ante Meridiem

American National Standard Institute.

Annunciator Response

Procedure

American Society. of Materials

'ureau

of Radiation Control

Computer Aids

Corrective Action

cubic centimeter per hr.

Component Cooling Water

Code 'of Federal

Regulations

CN

CHN

CR

D.C.

dpm

e.g.

EP IP

ERT

etc

oF

FL

FNE

FPL

GL

GN

GNN

gpm

HEPA

HHSI

HP

HPS

HPSS

HPT

Hx

Idm C

i.e.

IFI

IST

L

LER

NG

HCA

NDA

HM

HOV

HSSV

NCC