Text

Insp Repts 50-335/89-02 & 50-389/89-02 on 890123-27. Violations Noted.Major Areas Inspected:Licensee Conformance to Reg Guide 1.97,Rev 3
	ML17222A730
Person / Time
Site:	Saint Lucie
Issue date:	02/27/1989
From:	Conlon T, Fillion P; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML17222A728	List: ML17222A727; ML17222A729; ML17222A730;
References
	RTR-REGGD-01.097, RTR-REGGD-1.097 50-335-89-02, 50-335-89-2, 50-389-89-02, 50-389-89-2, NUDOCS 8903240279
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UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

101 MARIETTAST., N.W.

ATLANTA,GEORGIA 30323

Report Nos.:

50-335/89-02

and 50-389/89-02

Licensee:

Florida Power and Light Company

9250 West Flagler Street

Miami, FL

33102

Docket Nos.:

50-335

and 50-389

Facility Name:

St. Lucie

1 and

2

Inspection

Conducted:

J~a uary 23-27,

1989

Inspector:

P.

F. Fillion

Team Members:

M. N. Miller

C. J.

Paulk

License

Nos.:

DPR-67 and NPF-16

= "vg

Date Signed

Approved by:

T.

E. Conlon, Chief

Plant Systems

Section

Engineering

Branch

Division of Reactor Safety

Date Signed

SUMMARY

Scope

This

special

announced

inspection

was

conducted

to verify licensee's

conformance

to Regulatory

Guide 1.97,

Instrumentation for Light-Water-Cooled

Nuclear

Power

Plants

to Assess

Plant

and

Environs

Conditions

During

and

Following an Accident.

Results

The licensee

has

performed the installation

and modification of instrumentation

to

comply with Regulatory

Guide 1.97,

Revision 3.

However,

there

were

exceptions

noted

in the Safety

Evaluation

Report

dated

July 29,

1986,

and

subsequent

correspondence

with the

NRC.

Additional

items,

including

one

violation, were identified during this inspection

which are

discussed

in the

report.

Within the areas

inspected,

the following violation was identified.

Failure

to calibrate

six

RG 1.97

instrumentation

channels

and

loops,

paragraph

2.c.

8903240279

890303

PDR

ADOCK 05000335

6

One unresolved

item was identified involving the need to upgrade

EOP training

of operators relating to

RG 1.97 instrumentation,

paragraph

2.c.

Meakness

Several

discrepancies

and concerns

were identified which are discussed

in the

report.

These discrepancies

and concerns

appear

to be caused

by the licensee's

inattention to details.

In addition, it appears

that operations

personnel

did

not fully understand

the significance of

RG 1.97 instrumentation for post-

accident

monitoring.

This apparent

weakness

is

due to lack of training in

post-accident

monitoring of RG 1.97 instrumentation.

REPORT

DETAILS

Persons

Contacted

Licensee

Employees

G. Boi.ssy, Plant Manager

P. FitzRoy,

IKC Engineering

(JPN)

J. Harper, guality Assurance

Superintendent

S. Koleff, I8C Supervisor

C. Leppla,

IKC Superintendent

G. Maddem,

Nuclear Licensing

L. McLaughlin, Technical Staff Licensing

K. Mohindroo, Electrical/ISC Supervisor

(JPN)

L. Neely, Technical Staff Licensing

D. Parker,

I8C Engineering

(JPN)

G. Regal,

Maintenance

Management

Supervisor

D. Smith, Electrical/ISC Engineer

(JPN)

W. Windecker, Assistant Superintendent

- Procedures

D. Wolf, I8C Supervisor

Other

licensee

employees

contacted

during this

inspection

included

engineers,

operators,

technicians,

and administrative personnel.

Other Organization

S. D'Auria,

EBASCO

R. Gonzalez,

EBASCO

NRC Resident

Inspector

G. Paulk

Attended exit interview

Acronyms

and initialisms

used

throughout this report

are listed in the

last paragraph.

Inspection

of Licensee's

Implementation

of Multiplant Action A-17:

Instrumentation

for Nuclear

Power

Plants

to Assess

Plant

and

Environs

Conditions

During

and

Following

an

Accident

(Regulatory

Guide 1.97)

(25587)

Criterion

13,

"Instrumentation

and

Control," of Appendix

A to

10 CFR Part 50 includes

a requirement that instrumentation

be provided to monitor

variables

and

systems

over their anticipated

ranges for accident condi-

tions

as

appropriate

to ensure

adequate

safety.

Regulatory

Guide 1.97

(RG 1.97)

describes

a method

acceptable

to the

NRC staff for complying

with the Commission's

regulations

to provide instrumentation

to monitor

plant variables

and systems

during and following an accident.

The

purpose

of this inspection

was to verify that the licensee

has

an

instrumentation

system for assessing

variables

and

systems

during

and

following an accident,

as discussed

in Regulatory Guide (RG) 1.97.

Under

accident

conditions it is

necessary

that

the operating

personnel

have

(I) information that permits

the operators

to take

preplanned

actions

to

accomplish

a

safe

plant

shutdown;

(2)

determine

whether

the reactor

trip, Engineered

Safety-Feature

Systems

(ESFS),

and that other manually

initiated safety systems

important to safety are performing their intended

functions;

and (3) provide information to operators

that will enable

them

to determine

the potential

for causing

a

gross

breach

of .the barriers

to radioactivity release

and to determine if a gross

breach of a barrier

has

occurred.

For this

reason

multiple instruments

with overlapping

ranges

may

be necessary.

The required instrumentation

must

be capable of

surviving. the accident

environment for the length of time its operability

is required.

It is desirable

that components

continue to function follow-

ing seismic events.

As

a result, five types of variables

have

been specified that serve

as

guides

in defining criteria

and

the selection

of accident-monitoring

instrumentation.

The types

are:

Type A - Those variables that provide

information needed

to permit the control

room operating

personnel

to take

specified

manual

actions for which

no automatic control is provided

and

that are

required for safety

systems

to accomplish their functions for

design

basis

accident

events.

Type

B -

Those variables

that provide

information to indicate whether plant safety functions are

being

accom-

plished.

Type

C - Those variables

that provide information to indicate

the potential for barriers

being breached

or the actual

breach of barriers

to fission

product

release.

Type

D -

Those

variables

that

provide

information to indicate the operation of individual safety

systems

and

other systems

important to safety.

Type

E - Those 'variables

to be monitored

in determining

the magnitude of the release

of radioactive materials

and

for continuously assessing

such release.

The design

and qualification criteria are

separated

into three

separate

Category

1 provides

the most stringent

requirements

and is intended for

key variables.

Category

2 provides

less

stringent

requirements

and

'generally

applies

to 'instrumentation

designated

for indicating

systems

operating

status.

Category

3 is intended

to provide requirements

that

wi 11

ensure

that high-quality off-the-shelf instrumentation

is obtained

and applies

to backup

and diagnostic

instrumentation.

A key variable is

that single variable (or minimum number of variables) that most directly

indicates

the accomplishment

of a safety function (in the case of Type D)

or radioactive material

release

(in the case of Type E).

Type

A variables

are plant specific

and

depends

on the operations

that

the

designer

chooses

for planned

manual

actions.

Inspection

of

Category

2 Instrumentation

The instrumentation

listed in the Category

2 Tables of this section

was

examined

to verify that the design

and qualification criteria of

RG 1.97

had

been satisfied.

The instrumentation

was

inspected

by

reviewing drawings,

procedures,

data sheets,

Categories

1

and

2 instruments

are properly identified.

Addition-

ally, the licensee

stated

that certain

Dl and

2 instruments

may

be

identified and that the marking would be consistent

between units.

This project for the identification of instruments

is for both uni ts

and

the simulator.

The simulator is included

as

the result of

a

walkdown which identified inconsistencies

between

the plant and the

simulator.

14

During the review of the

Type

A variable, it was noted that the

EOPs

required

the

operator

to verify reactor

power less

than

5X and

decreasing.

If this condition was not met,

the operator

was required

to take

manual

actions

in order to maintain

the plant in

a safe

condition or to mitigate the accident.

The licensee

had not included

neutron flux in the Type

A variables,

but will include it when the

list is revised.

Also during the review of the

EOPs for use of

RG 1.97 instrumenta-

tion, the inspector

conducted table top walkthroughs of the

EOPs with

a licensed

operator.

The inspector

found that the operator did not

refer to the

RG 1.97 instrumentation

unless it was the instrumenta-

tion normally

used

for day-to-day

operations.

Further inspection

identified this to be the

norm for operations.

The training program

did not stress

the necessity of using

RG 1.97 instrumentation

and,

as

such, it wou'Id

be possib'le for the operators

to rely on inaccurate

and/or failed instruments.

The licensee

committed to add additional

words to an existing note in the

EGPs to advise the operators

to use

the

RG 1.97 instrumentation

during accident

conditions.

Also, the

licensee

committed to enhance

the simulator training to emphasize

the

RG 1.97

instrumentation.

This is identified

as

Unresolved

Item

50-335, 389/89-02-02,

Upgrade of EOP Training for Operators.

A comparison

of the

RG 1.97 variables

selected

with the

EQ Master

List did not 'result in any discrepancies

being identified.

Those

RG 1.97 components

that were required to be environmentally qualified

were found to be

on the

EQ Master List.

The results of previous

EQ

inspections

are

documented

in

NRC

Report

Nos. 50-335/86-08,

50-389/86-07,

50-335/88-24,

50-389/88-24.

The licensee

provided

a list of modifications that were

performed

after the Safety Evaluation

Report

was issued in July 1986.

A total

of seven

Plant Change/Modification

packages

were required to perform

the modifications.

Four of the packages

were reviewed

and

found to

be acceptable

to ensure

compliance with

RG 1.97.

These

packages

were

for the

Excore Neutron

Flux Monitoring System

(PCM 228-184);

the SIT

and

Containment

Fan

Cooler Instrumentation

Upgrade

(PCM 128-187 for

Unit I and

PCM 129-287 for Unit 2); and,

the

EQ List revision that

re-analyzed for a mild environment

(PCM 77-186).

All,of the modifi-

cations

have

been

completed with the exception of

PCM 129-287 which

. is scheduled for the upcoming outage.

The licensee,

in response

to the announced

RG 1.97 inspection,

pulled

and

reviewed

the

RG 1.97

instrumentation

calibration

data

sheets.

During the

review for Unit I, the

licensee

discovered

that six

RG 1.97 instrumentation

channels

and flow loops

were not calibrated

within the

time

frame

required

by

the calibration

Procedure

No. 1-1240065

and

the

preventative

maintenance

program.

The

instrumentation

channels

and loops are

two Category I Type A Neutron

Flux

channels

and

four Category

2

Type

D High Pressure

Safety

Injection (HPSI) flow loops.

15

The

Neutron

Flux channels,

identified by reactivity elements,

are

RE-26-80A

and

RE-26-80B.

The licensee

stated

these

channels

were

installed

as

a

RG 1.97 modification and initially calibrated

during

December

1985.

18C

Procedure

No. 1-1240065,

Excore

Neutron

Flux

Monitor quarterly,

was written

and

approved

May 3,

1986.

This

procedure

requires

a quarterly calibration.

The licensee calibrated

both

channels

in June

1986 which was the last time

Channel

A was

calibrated.

Channel

B

was

only calibrated

one

more

time,

February 25,

1988.

The licensee

stated

both channels

were inadver-

tently

removed

from the

preventative

maintenance

tracking

system

which schedules

calibrations.

The

four

HPSI

flow loops,

identified

by flow transmitter,

are

FT-3313,

FT-3323,

FT-3333,

and

FT-3343.

These

HPSI flow loops were

last calibrated

November 23,

1985.

The licensee

did not

have

a

specific procedure requiring calibration

and did not have

them listed

in

the

preventative

maintenance

tracking

system

for scheduled

calibration.

The licensee

stated

these

HPSI flow loops should

be

on

an eighteen-month

calibration schedule.

The failure of the

licensee

to calibrate

the important to safety

RG 1.97 Neutron Flux Channels

A 5

B and the four HPSI flow loops is

identified's

a violation.

This violation includes

the failure of

the licensee

to have

a specific procedure

requiring calibration for

the

HPSI flow loops, failure to calibrate

the Neutron Flux Channels

A

5

B as

required

by the calibration

procedure,

and

not having this

instrumentation

in the preventative

maintenance

tracking system for

scheduling

periodic calibration.

This violation is identified

as

50-335/89-02-01,

Failure to Calibrate

RG 1.97 Instrumentation

and is

applicable to Unit I only.

Violation 50-335/89-02-01

is not considered

as licensee-identified.

The

licensee

initiated the review of the calibration

data

sheets

after being

informed during the first week of December

1988 that the

NRC would perform the

RG 1.97 inspection

during January

1989.

In

addition,

the licensee

was

informed that calibration

data

sheets

would be

reviewed

by the

NRC to verify the

RG 1.97 instrumentation

was calibrated.

Corrective Action Program

The licensee

took corrective action to calibrate

the six instrumen-

tation

channels

and flow loops.

Plant

Work Order

No.

6493

dated

January

17,

1989,

was issued

to calibrate the Neutron Flux Channels

A

and

B.

Plant

Work Order

No.

6507 dated January

19,

1989,

was issued

to calibrate

the four HPSI flow loops.

The inspectors verified that

the Neutron

Flux Channels

A 5

B and the four HPSI flow loops were

calibrated satisfactorily.

16

However,

the inspectors

were

concerned

that other instrumentation

important to safety

may not have specific calibration procedures

and.

may not

be

in the

preventative

maintenance

tracking

system for

scheduled

calibration.

The

licensee

stated

that

guality

Assurance

has

performed audits to verify that instrumentation will be

properly calibrated

and will continue to perform audits for instru-

mentation calibration requirements.

3.

Exit Interview

The inspection

scope

and results

were summarized

on January

27,

1989, with

those

persons

indicated

in paragraph

1.

The inspectors

described

the

areas

inspected

and

discussed

in detai I the inspection results listed

below.

Proprietary

information

is

not

contained

in this

report.

Dissenting

comments

were not received

from the licensee.

However,

the

licensee

did

make

several

oral

commitments

which

are

discussed

in

paragraph

2.c.

One violation was identified,

50-335/89-02-01,

Failure to calibrate six

RG 1.97 instrumentation

channels

and loops,

(paragraph 2.c.).

One unresolved

item was identified, 50-335,

389/89-02-02,

Upgrade of

EOP

Training for Operators.

'17

and Initialisms

AN

Eg

FI

FIA

FPL

FR

FT

HPS

LI

LIA

LIC

LIS

LR

LT

PCN

PI

PIA

PR

RE

RI

TT

TY

VM

Auxiliary Feedwater

Analytical Indicating Instrument

Ammeter

Emergency Operating

Procedure

Environmental qualification

Flow Indicator

Flow Indicating

Florida Power

5 Light Company

Flow Recorder

Flow Transmitter

High Pressure

Safety Injection

Level Indicator

Level Indicating

Level Indicating Controller

Level Indicating Switch

Level Recorder

Level Transmitter

Plant

Change Notice

Pressure

Indicator

Pressure

Indicating

Pressure

Recorder

Pressure

Transmitter

Request Additional Information

Reactor Coolant System

Reactivity Element

Regulatory Guide

Reactivity Indicator

Reactivity Recorder

Safety Evaluation Report

Safety Injection Tank

Temperature

Element

Temperature

Recorder

Temperature

Transmitter

Temperature

Signal Conditioner

Voltmeter

ML17222A730

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