ML18005A683

instrumentation.

This

drawing

needs

to

be

updated.

The

licensee

stated that this would be accomplished.

8811030356

881025

PDR

ADOCK 05000400

6

PNU

A weakness

was

identified

in

the

licensee's

program

concerning

trending

and

evaluation

of as-found

out of tolerance

reading

on

instrument calibration (Paragraph 2.c.(2)(a)).

REPORT DETAILS

1.

Persons

Contacted

Licensee

Employees

J. Brown, Senior Specialist,

Corporate

QA

W. Edwards,

I&C Engineer,

Nuclear Engineering

Dept.

E. Evans,

I&C Project Engineer,

Nuclear Engineering

Dept.

"C. Hinnant, Plant General

Manager

"A. Howe, Senior Specialist,

Regulatory Compliance

"M. Jackson,

Electrical/I&C Supervisor

D. McCarthy, Principal Engineer,

Nuclear Engineering

Dept.

"C. McKenzie, Principal

Engineer,

Quality Assurance/Quality

Control

"T. Morton, Manager of Maintenance

J.

Presson,

Specialist,

Nuclear Engineering

Dept.

K. Russell,

On-site Nuclear Safety

"M. Turkal, Senior Specialist,

Nuclear

Fuel

& Licensing Dept.

"R.

Van Metre,

Manager of Technical

Support

Other

licensee

employees

contacted

during

this

inspection

included

engineers,

operators,

mechanics,

security force members,

technicians,

and

administrative

personnel.

NRC Resident

Inspectors

W. Bradford, Senior Resident

Inspector

"M. Shannon,

Resident

Inspector

Attended exit interview

2.

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

conditions

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

Commissions

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

RG 1.97.

Under accident conditions

it is

necessary

that

the operating

personnel

have

(1) information that

permits the operator

to take preplanned

actions to accomplish

a safe plant

shutdown;

(2) determine

whether

the

reactor

trip,

Engineered

Safety-

Feature

Systems

(ESFS),

and manually initiated safety

and other

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 radioactive

release

and to determine if a breach

has occurred.

It is essential

that

indicators

be

provided

to

the

operator if the

barriers

are

being

challenged that will allow the release

of radioactive materials.

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

components

continue to function following 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

8 -

Those

variables

that

provide

information

to

indicate

whether

plant

safety

functions

are

being

accomplished.

Type

C -

Those

variables

that

provide

information

to

indicate the potential for barriers

being breached

or the actual

breach of

bar riers

to fission

product

release.

Type

D -

Those

variables

that

provide information to indicate 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

system

operating

status.

Category

3 is

intended

to provide

requirements

that

will ensure

that high-quality off-the-shelf instrumentation

is

obtained

and applies

to backup

and diagnostic

instrumentation.

A key var'.able is

that single

accomplishment

of

a safety function

(Types

8

and

C), or the

operation

of

a

safety

system

(Type 0),

or radioactive

material

release

(Type E).

Type

A variables

are

plant

specific

and

depends

on

the

operations

that

the

designer

chooses

for

planned

manual

actions.

Inspection

of Categories

1

and

2

equipment,

was

performed

as

described

below.

Category

1 Instrumentation

The instrumentation listed in the Category

1 Table, of this section,

was examined

to verify that the design

and qualification criteria of

had

been

satisfied.

The

instrumentation

was

inspected

by

reviewing drawings,

procedures,

data

sheets,

Category

2 Instrumentation

The instrumentation

listed in the Category

2 Table, of this section,

was examined to verify that the design

and qualification criteria of

had

been satisfied'he

instrumentation

was

inspected

by

reviewing drawings,

procedures,

data

sheets,

Category

1 variable.

In addition,

RG 1.97 specifies

these

variables

as

Type

B, Category

1.

RG 1.97 specifies

that

Category

1

variables

should

be

monitored

by

redundant

instrument loops.

The Harris plant has

one instrument

loop

per reactor

loop for the

Th variable.

However,

the

power

supply for each

loop is the Channel

I inverter.

Therefore,

11

the Harris plant

does

not

have

redundant

instrumentation

for Th.

The

Tc variable

has similar instrumentation

except

that it is

powered

from the

Channel

II inverter.

The

licensee

states

in his submittal,

dated

September

6,

1983,

that diverse

information for Th is provided

by the

core

exit

thermocouples

(Type

C,

Category 1),

and

diverse

information for Tc is provided

by the

steam line pressure

(Type

A, Category 1).

Nevert'heless,

the lack of redundant

power supplies for the

Th and

Tc variables

may constitute

a

deviation

from

RG 1.97.

Since

the

licensee's

submittal did not clearly describe

the

power supplies,

this

matter

was

not specifically reviewed

by the

NRC.

in order

to resolve

the

matter,

the

licensee

agreed

to

make

a

supplementary

submittal

to

NRC describing

their

instrumentation

for

Th

and

Tc,

diverse

measurements

instrumentation

power supply

and

any justification they

may

wish to provide for acceptability

of the present

design.

This

item

is designated

as

URI 50-400/88-29-02

pending

licensee

and

NRR action

on this item.

(Open)

50-400/88-29-02,

Update

and

Licensee's

Drawing for

PAM Instruments,

and

Ensure

MCB

has

Proper

Designator

for

PAM Instrument.

A review of

the

PAM

instruments

in

the

and

those

listed

on

Drawing

No. 2166-S-9000,

Rev. 0,

Post

Accident

Monitoring

Equipment,

identified

several

discrepancies.

Some

PAM

items that were listed in the

FSAR were not listed

on the

drawing

and

some

PAM items that were listed in the drawing

were not in the

FSAR.

In addition,

Note

1 in the drawing

did not provide

a complete definition of all variables

and

the

drawing

was

not

annotated

to

show

the

instrument

indicators

that

had

the

common

PAM designator

on

the

control

panel

(the later is considered

to be

an enhancement

for the

drawing).

The

licensee

committed to review

and

revise both of these

documents

such that the next update of

the

(which will be approximately

October

1989) will

include the changes.

RG 1.97,

Rev.

3,

requires

that

"Types

A,

B,

and

C

instruments

designated

as

Categories

1

and

2

should

be

specifically identified with

a

common designation

on

the

control panels

so that the operator

can easily discern that

they

are

intended

for

use

under

accident

conditions."

During inspection

in the Control

Room,

discrepancies

were

noted

in the distinct labelling of

PAM instruments.

For

example,

all

Type A, Category

1 variables

did not

have

a

common designator

(a yellow border trimming the

PAM readout

indicator)

on the control

board.

Some

Type

D, Category

2

variables did.

Type

D PAM instruments

are not required

by

12

R.G.1.97

to

have

this distinct labeling.

To

add

these

common designators

for Type

D instruments

is considered

to

be

a licensee's

prerogative.

The licensee

has committed to

review the control

board designations

to ensure

that all

Type A,

B,

and

C

variables

of Categories

1

and

2

are

properly

labeled.

The

licensee

also

stated

that

an

evaluation will be

performed

for those

Type

D variables

that

are

designated

for accident

conditions.

This

is

identified

as

50-400/88-29-02,

Update

FSAR,

PAM

Instrument

Drawing and

MCB PAM Designations.

Control

Room Envelope

states

that it provides

the

minimum

number

of

variables

to

be

monitored

by

control

room

operating

personnel

during and following an accident.

This statement

could

be interpreted

to

mean that the variable

should

be

monitored in the control

room.

The

1'icensee

has

designated

containment

concentration

as

a

Type A, Category

1 variable.

This is

not

monitored

in

the

control

room.

The

licensee's

submittal,

dated

September

6,

1983, indicates that percent

concentration

is displayed

in the

control

room

envelope,

processed

by the

computer

and

recorded.

During

the

inspection, it was

confirmed that

percent

concentration

in

the

containment

is

displayed

on

a

continuous

real

time indicator and continuously recorded at

the

analyzer

panel.

The

panel

is located

in

a

room adjacent

to

the

main

control

room

that

would

be

readily accessible

during

and following an accident.

The

variable is scanned

by the plant computer at

an appropriate

interval

and

may be displayed

on demand

on

a

CRT screen

in

the

main control

room.

The variable is also

recorded

by

the computer.

Based

on the

above

and

NRR's acceptance

of

the

licensee

submittal,

the

instrumentation

provided

(indicators

and recorders)

to monitor containment

concentration

in the control

room envelope is acceptable.

Concerns

Identified

By a Previous

Inspection and/or During

Part of the Malkdown Inspection for

PAM Equipment

During

a

routine

gA inspection

in October

19-23,

1987,

unterminated

cables

were

observed

in

the

relay

and

auxiliary relay

panels

outside

the control

room.

These

unterminated

cables

were

each

tagged

with

a

cable

identification

number which was identical

to

an installed

cabl'e

in the

same

panel.

The

licensee

researched

this

problem

and

determined

that the unterminated

cables

were

"spared"

and

should

be identified

as

abandoned

or spared

cables.

13

The

Site

Engineering

Unit

(SEU)

wrote

PCR-2444

dated

October 26,

1987

and

Form

2 dated

November 5,

1987,

to

specify

the

corrective

action

for the

spared

cables.

Technical

Support

issued

Form

2ÃX to close

PCR-2444

and

initiated

work request

authorization

WRA-87B1MF1.

This

work

ticket

was

cancelled

without

the

work

being

accomplished.

The inspector

was

informed that

WRAs

have

been

and

can

be

cancelled

without

the

work

being

accomplished

as

required.

This concern will be

addressed

by the resident

inspectors.

The

inspector

examined

relay

cabinet

ARP-4B/SB

and

identified fuses

Ll-2609 and

L5-2640

as

the Fault Trap

FT

6R type which did not appear to properly fit into the fuse

holders.

The licensee's

engineering

personnel

could

not

satisfactorily

explain this condition at the time.

This

concern will be addressed

by the resident

inspectors.

3.

Action On Previous

Inspection

Findings (92701)

(Closed)

IFI

50-400/86-42-01,

Review

Testing

Instructions

for

Periodic

Testing

of Circuits

Covered

by Appendix

R Coordination

Study.

In

a letter

to the

NRC,

dated

August 6,

1986,

(Serial:

NLS-86-270)

CP6L described

a program for periodic testing of circuit

breakers.

The

purpose

of this

program is to demonstrate

that

the

breaker

coordination

study performed pursuant to

NUREG 0800,

Section

9.5. 1,

Fire Protection

Program,

remains

valid; i.e.,

the

breaker

time-current characteristics

have

not significantly drifted.

At the

time of the fire protection

program inspection

conducted

on June 3-6,

1986,

the

program for periodic testing of circuit breakers

was not

yet in place.

An Inspector

Follow-up Item was established

to ensure

NRC

review of the

program.

That review was

conducted

during this

inspection (88-29),

and the program

was found to be acceptable.

In general,

ten percent

of each

type of circuit breaker

is tested

during

each

refueling

outage.

The Planning

Section

issues

a work

request for the testing of circuit breakers.

A maintenance

engineer

maintains

a list of breakers

to-be tested

and

records

of testing

dates

for each.

By referring to the

master list,

a

ten

percent

sample list can

be developed

for

a particular

outag'e.

The

sample

list is attached

to the work request

and given to the

foreman for

testing.

Testing is performed according to the following procedures,

which were reviewed

by the inspector:

Procedure

No:

MPT-E0023,

Revision 0,

480

VAC Power Circuit

Breaker (Safe

Shutdown) Solid State Trip Unit Test

This procedure

covered

long time element pick-up and time delay;

instantaneous

pick-up;

and short-time pick up and time delay.

14

Procedure

No:

MPT -

E 0022,

Revision 0, General

Electric 6.9

KV

Overcurrent

IFC-53 (Safe

Shutdown)

Relay Calibration

This procedure

covered pick-up and target test,

time delay test

and instantaneous

pick-up test.

Procedures

No.:

MPT E0024,

Revision 0, Molded - Case Circuit

Breakers

(Safe

Shutdown) Test.

This procedure

covered inspection

and maintenance,

thermal trip

,test

and instantaneous

trip test.

This item is closed.

(Closed)

URI 50-400/86-62-02,

Review the Reportability of all NCR's

Involving Electrical

Cable

Separation.

Inspection

at

the

site

conducted

during

the

construction

phase

in the

area

of electrical

separation

resulted

in an Unresolved

Item being identified.

Numerous

Nonconformance

Reports

for separation

criteria violations

had

been

generated

by the

licensee.

However,

at the

time of the inspection

(July 7-25,

1986),

the

'nonconformances

were

not

being

reported

pursuant to 50.55(e).

The inspector believed

the separation criteria

nonconformances

were

reportable

because

they were discovered

after

the work was signed-off

as being completed

and inspected.

Subsequent

to

the

inspection,

the

licensee

did report electrical

separation

criteria nonconformances

pursuant

to 50.55(e).

The final report

on

the

separation

issue

was transmitted

on December

12,

1986,

by letter

No. HO-860395(o).

During this

inspection

(88-29),

the

licensee's

internal

reports

to resolve

the

issue

were

reviewed.

Corrective

Action/Noncomformance

Report 86-0517

was closed

on October

13,

1986.

This item is closed.

(Closed)

URI 50-400/86-88-02,

Evaluate

Repor tabi lity of Inadequate

Preop

Testing

of

ESF

Components.

It

was

establ i shed

during

an

inspection

conducted

on

November

17-21,

1986,

that

inadequate

preoperational

testing

of

the

sequencer

panel

constituted

a

viol,ation.

The

violation

was

issued

on

January

8,

1987.

The

licensee

responded

to

the violation,

and

the

response

has

been

reviewed

and

accepted

by the

NRC.

Unresolved

Item 86-88-02

was

associated

with this violation and concerned

the reportability of the

inadequate

preoperational

test

and

a defective

sequencer

panel.

It

has

been

determined

by the

NRC that

the

inadequate

preoperational

test procedures

was not reportable

under 50.55(e)

because it did not

constitute

a significant

breakdown

of

an entire

portion of the

quality assurance

program itself.

The defective

sequencer

panel

was

not reportable

under Part

21 because

the deficiency was already

known

to the

NRC and did not constitute

a possible

generic

problem.

This item is closed.

15

d.

(Closed)

BU-88-03,

Inadequate

Latch

Engagement

in

HFA Type Latching

Relays Manufactured

by G.E.

IEB 88-03 was issued

March 10,

1988,

and

requested

that licensees

ensure

that all

GE latching-type

HFA relays

installed in Class

lE applications

have adequate

latch engagement

and

that those relays which fail to meet acceptance

criteria

be repaired

or replaced.

By letter dated

June

22,

1988, to NRC,

CP5L stated that

there were

no

HFA type latching relays of the series

identified in

the bulletin in use

in safety-related

applications

at

SHNPP.

This

was also

documented

in the licensee's

response

to

IEB 84-02,

HFA

Relays

in Class

1E Safety Systems.

Based

on the above, this item is

closed.

4.

Exit Interview

The

inspection

scope

and results

were

summarized

on

September

2,

1988,

with those

persons

indicated in paragraph

1.

The inspectors

described

the

areas

inspected

and discussed

in detail

the inspection

items

and concerns.

Proprietary information is not contained

in this report.

In the areas

inspected,

one deviation

and

two unresolved

items

'"(URI) were

identified as indicated below.

Unresolved

items are matters

about which more information is required

to determine

whether

they are

acceptable

or

may involve violations

or deviations.

Deviation

50-400/88-29-03,

Program Deficient for Trending

and Evaluating

As-found

Out-of-Tolerance

Readings

During

Instrument

Calibrations.

Paragraph

2.c.(2)a.

50-400/88-29-01,

Licensee

to Submit Written Justification

on

RCS

Temperature

Elements

Concerning

Redundancy.

Paragraph

2.c(2)b.

50-400/88-29-02,

Update of Licensee'

Drawing 2166-S-9000.

Paragraph

2.c.(2)c.

5.

Acronyms, and Initialisms

AUX FW (AFW)

BU

CRT

OEV

DPT,

EQ

FI

FT

IFI

LI

Auxiliary Feedwater

(System)

NRC Bulletin

Cathode

Ray Tube

Deviation

Differential Pressure

Transmitter

Environmental Qualification

Flow Indicator

Flow Transmitter

Final Safety Analysis Report

Inspector

Followup Item

Level Indicators

16

LIR

LR

LT

MCB

PR

RC

SEU

SHNPP

TE'I

TIR

TR

TT