ML17342A647

loads

beyond

those

analyzed

in

JPE-L-86-59,

Revi sion

1;

see

paragraph

9.

No Notice of Violation for this item will be

included in the

report as this matter is being considered for enforcement

action

as part of

a separate

report.

8608110623

860714

PDR

ADOCK 05000250

8

PDR

'1

r

X

'4

0

REPORT DETAILS

1.

Persons

Contacted

Licensee

Employees

¹C. H. Methy, Site Vice President

~ C. Baker, Plant Manager (Nuclear)

"¹D. Grandage,

Operations

Superintendent

(Nuclear)

¹W. Hiller, Training Superintendent

"¹V. Kaminskas,

Operations

Supervisor

'¹J. Arias, Regulations

and Compliance Supervisor

¹D. Jones,

Procedure

Upgrade

Program

(PUP) Supervisor

¹J. Strong, Electrical Supervisor

Other licensee

employees

contacted

included engineers,

operators,

and office

personnel.

NRC Resident

Inspectors

t

~¹T. Peebles

R. Brewer

~Attended exit interview on April 6,

1986.

¹Attended exit interview on April 18,

1986.

2.

Exit Interview

The inspection

scope

and findings were

summarized

on April 6 and 18,

1986,

with those

per sons indicated in paragraph

1 above.

The inspector

described

the

areas

inspected

and discussed

in detai

1 the inspection

findings.

No

dissenting

comments

were received

from the licensee.

The licensee

did not

identify as proprietary

any of the materials

provided to or reviewed

by the

inspector during this inspection.

The licensee

made

the following commit-

ments at the exit interview:

a.

To revise

emergency

operating

procedure

(EOP) attachments

as necessary

to resolve identified deficiencies

which

have

the potential

to allow

overloading the emergency diesel

generators

(EDG's).

b.

To revise the

EOPs to reflect

a maximum

EDG loading of 2845

KM.

C.

To revise training brief 122 on

EDG loading precautions

to reflect the

significant changes

due to Revision

1 to JPE-L-86-59.

The revisions

to this training brief were to be provided to all licensed

personnel.

To

provide

locking

mechanisms

for intake

cooling

water/component

cooling water

(ICW/CCW) heat

exchanger

isolation valves throttled to

ensure that combined

ICW/CCW pump

EDG loads are maintained

below 500

KW.

t

ll

e.

To provide additional training to licensed

personnel

in the following

areas:

Integrated use'f

the

EOP's

and

attachments

loading.

This training

was to include in-plant walkthroughs of

the

procedures

and

attachments

for

each

licensed

operator.

Licensed

individuals participating in,accelerated

requalification

training were

to receive

the additional

EOP training prior to

resuming shift responsibilities.

Placing

the, control

room chillers in service

as required

by the

EOP's.

Gamma

Metrics

post accident flux monitoring instrumentation

for

licensee

personnel

from Hot License

Class

10.

This training was

to be conducted

immediately

as licensed

personnel

from this class

had

primary

responsibility

for

operation

of

Unit 3

pending

completion of additional

NRC requalification examinations.

3.

Licensee Action on Previous

Enforcement Matters

(Closed)

UNR (250,251/85-22-02):

The

licensee

was

unable

to retrieve

QA

records relating to a specific plant work order

(PWO) maintenance activity.

The inspector

reviewed several

PWOs and requested

the clearances

associated

with these

PWOs.

The licensee

was able to retrieve all clearances

requested.

The inability to retrieve

the

clearance

associated

with this

UNR is

considered

an isolated incident.

This

UNR is closed.

(Closed) Violation (250,251/85-22-03):

Item a.

The licensee

failed to establish

adequate

maintenance

procedures

to ensure

the proper wiring of the

DC input filter circuit of 4A

static inverter

which resulted

in the

mi swiring of the filter

circuit and contributed to reactor trips on September

20,

1984 and

October 9,

1984.

Item b.

The licensee failed to implement Administrative Procedure

0190. 19,

Control

of

Maintenance

on

Nuclear

Safety

Fire

Protection

Systems,

in the rewiring of the input filter section of

the

4A inverter.

This rewiring was

performed

under

a

PWO which

did not define the work to be done or any

QC inspections

or hold

points.

e

The inspector

reviewed revised

procedure

AP 0190. 19,

Control of

Maintenance

on

Safety

Quality

revision dated January

8,

1986,

O-ADM-701,

PWO Preparation,

revision

dated

March 25,

1986,

and several

inter-office correspondences

on

administrative

guidelines

to

nuclear

electrical

maintenance

personnel.

These

procedures

and electrical

department

guidelines

provide

more

detailed

instructions,

and

increase

the

use

of

supervisory

holds

in preparing

PWO work descriptions.

Violation

examples

a.

and b.

are

adequately

addressed,

and

these

examples

are closed.

The licensee failed to establish

abnormal

operating

procedures

to

contend

with the

loss

of the

4A motor control

center.

This

resulted

in the

4AA05 and

4AB05 bus

supply fans

being

rendered

inoperable

due

to operators

failing to close

breaker

40521

on

May 17,

1985.

The inspector

reviewed 3/4-0P-007,

480 Volt Motor Control Centers,

revision dated August 30,

1985.

The procedures

have

been

revised

to require

the operator

to verify that all motor control center

breakers

are reset.

Additionally, this

requirement

is

indepen-

dently verified.

Violation example c. is closed.

The licensee

failed to implement Administrative Procedure

0103.3,

Use of Temporary

System Alterations (TSA),

on July 6,

1984, for a

temporary

system

alteration

to the

3C Accumulator hi-low level

circuit, in that,

the licensee failed to maintain documentation

of

the TSA.

The

inspector

reviewed

a revision,

dated

January

14,

1986,

to

O-ADM-503, Control

and

Use of Temporary

System Alterations (the

replacement

procedure

for AP-0103.3).

0-ADM-503 requires

that

a

TSA log

be

retained

as

a quality assurance

record.

Violation

example d. is closed.

The

licensee

failed to

implement

maintenance

procedure

9707.1,

Inverter Periodic Inspection.

The

licensee

informed

the

inspector

that the procedure

upgrade

program was currently revising maintenance

procedures

to make

them

easier

to

work with.

The

inspector

reviewed

numerous

new

procedures

to determine if they were easier

to follow.

These

new

procedures

require the journeyman to document actions,

and include

more

supervisory

hold

points

and

gC

inspection

points.

The

completion of the procedure

upgrade

program should result in more

usable

maintenance

procedures.

Violation example

e. is closed.

The licensee failed to implement Administrative Procedure

0190.26,

section 8.51,

and perform preventive

maintenance

procedure

PM-74035,

Calorimetric Instrumentation

Periodic Calibration, at the frequency

designated

by the computerized

preventive maintenance file.

The licensee

planned

to reduce

the

frequency of this calibration

to

an

annual

PM, but after reviewing the past history of the

PM,

the

frequency

was

not

changed.

The

inspector

reviewed

the

computerized

preventive

maintenance file for

PM-74035

and it had

been

completed.

Violation example f. is closed.

LJ

(Closed)

UNR (250,251/85-22-04):

All post accident

sampling

system

(PASS)

PMs were delayed

due to system modifications.

The inspector

reviewed

the

PASS

PMs and all

had

been

completed with the exception

of calibrating the

nuclear data

computer

channels for each

PM.

The inspector

was informed that

the

computer

channels

could

not

be

calibrated

unti 1

the

computer

was

reprogrammed

to accept

the

PASS inputs.

This

UNR will be closed out, but

the

completion

ot

the

PASS

PMs will

be

an

inspector

followup

item

(250,251/86-.24-01).

(Closed)

UNR

(250,251/85-22-12):

Provide

documentation

justifying the

installation

of other

than,

the

approved

breaker

specified

in

plant

change/modification

PC/M

80-31.

'The

inspector

reviewed

a letter

from

Westinghouse

Electric

Corporation

to

Turkey

Point

Plant,

Electrical

Department,

dated

June

28,

1985,

confirming

that

circuit

breaker

P/N 1250C29G04

is

the

same

as circuit breaker

P/N

1268C14G04.

This

UNR is closed.

Unresolved

Items"

One unresolved

item was identified during the inspection:

UNR (250, 251/86-24-03):

Utilization of operator

action statements

in

notes

(paragraph

6).

Reactive

Inspection

Background

Florida

Power

and

Light Juno

Plant Engineering

(JPE)

conducted

a Safety

Evaluation,

JPE-L-86-59,

Revision

1,

which indicated

special

restrictions

that

must

be

placed

on Unit 4 during

the

operation

of Unit 3.

These

restrictions

were necessary

to ensure that the diesel

generator

loads that

could

be

experienced

during

design

basis

events

do not exceed

technical

specification

or design

limitations.

JPE identified that

the

KW rating

assumed

in the

FSAR for the intake cooling water ( ICW) and component cooling

water

(CCW) pumps

was based

on

a design

point that could

be

exceeded

with

only one

ICW and

one

CCW pump in operation.

The evaluation indicated that

with ICW/CCW pump runout,

the

2?50

KW auto-connected

technical

specification

limit and the 2950

KW FSAR 168 hour0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months

EDG emergency rating could be exceeded.

An additional

evaluation

was

performed

to

provide

a Justification

for

Continued Operation

(JCO) for operation of a single unit with the other unit

in cold

shutdown.

The evaluation

concluded

that Unit 3 could

be operated

with Unit 4 in cold

shutdown provided that the loads which would be placed

on the diesel

generators

or

on

a single diesel

generator

in the event

one

diesel

is inoperable,

are

limited through

recommended

restrictions.

The

recommended

restrictions

included

general

areas

such

as

reduced

flows

through

CCW

and

ICW, locking out

unnecessary

electrical

loads

such

as

instrument air

compressors

and

Unit 4

normal

containment

coolers,

and

revising

the emergency

operating

procedures

(EOPs) to restrict the addition

of diesel

loads.

~ An Unresolved

Item is

a matter which more information is required to determine

whether it is acceptable

or may involve a violation or deviation.

~ I

,Ij

JPE-M-86-18,

which

was

submitted

with Revision

0 of JPE-L-86-59,

was

an

-evaluation of the Turkey Point

The

EOPs were rewritten to incorporate

the

recommendations

of

the

Westinghouse

Emergency

Response

Guidelines

(ERGs).

JPE-M-86-18 contains

a

10

CFR 50: 59 review of the revised

by

Westinghouse,

and also

a

JPE

review of the

based

on emergency diesel

generator

(EDG)

loading

concerns

under

LOCA conditions.

The

JPE

review

determined

the

proposed

.EOPs

to

be adequate,

assuming

compliance with the

recommendations

in the

JCO and attachments.

Reactive

inspections

were conducted

by Region II personnel

on April 3-6 and

16-18,

1986,

to verify compliance with the recommendations

and limitations

detailed in JPE-L-86-59, "Justification for Continued Operation with One Unit

at Power and

One Unit in Cold Shutdown Relating to Emergency

Diesel Generator

Loads",

Revision

1, April 8,

1986.

The

combined

objectives

of the

two

inspections

included the following:

Review the incorporation of recommendations

of JPE into Turkey Point

The basis for this review was Revision

0 of JPE-M-86-18.

Review the adequacy

of licensed operator training on the

EOP revisions

resulting

from the Westinghouse

recommendations

and the recommendations

resulting

from the JPE

EDG loading evaluation.

Review resolution of the recommendations,

which are considered

commit-

ments

to

the

NRC,

as

contained

in JPE-L-86-59,

Revision

1,

dated

April 8,

1986, or in related

correspondence

including:

(1)

Unit 4 valve

and

breaker

alignments

necessary

to ensure

adequate

EDG capacity with an accident

on Unit 3 and

a loss of one of the

two

Review

of this

area

included

the

established

administrative controls over these

valve and breaker positions,

as

well as

a plant walkdown of the actual

alignments.

(2)

KW load limitations established

by procedures,

operator aids

and

training, or valve and breaker alignments.

(3)

Compliance

with

Confirmation

of Action Letter

(CAL) 50-250,

251/86-01 which confirmed actions to be completed prior to restart

of Units

3 and 4.

(4)

The

10 CFR 50.59 evaluation

submitted

on

EDG load limitations as

an

attachment

to JPE-L-86-59,

Revision

1

and

in

response

to

CAL 86-01.

Wal kdown

and

veri fy the throttled

val ve

positions

establ i shed

by

Special

Test 86-05

on the

CCW system.

Verify the adequacy of the revised Unit 3 startup shift staffing.

Review

Group

I

upgrade

requalification

training

effectiveness

and

progress.

II

v l.

Emergency Operating

Procedures

(EOPs)

In response

to the

EDG overload potential,

and the recommendations

contained

in

JPE-M-86-18,

the

licensee

made

significant

revisions

to

the

associated

with

LOCA or

loss

of offsite

power conditions.

The

general

philosophy utilized

by

the

Procedures

Upgrade

Plogram

(PUP)

staff

in

revising the

EOPs to meet

JPE recommendations

was to place

added actions

and

cautions

in .attachments

to the

whenever feasible.

Since the licensee

considers

the present operational restrictions

to be

a temporary condition,

this method of revision prevents

altering

the flow of the

permanent

This additional

reliance

on

attachments

to the

appears

to

be

less

effective

and more cumbersome

than placing the necessary

additional actions

and

cautions

directly into

the

appropriate

procedure

sections.

This

methodology

can

work provided that

each

EOP affected

by

an

attachment

contains

a reference

to the attachment,

or

a restatement

of the action or

caution

as contained

in the related attachment.

The inspection

in this area

was therefore

focused first on ensuring

that

each

recommendation

in

JPE-L-86-59,

Revision

1,

and JPE-M-86-18

was incorporated into the appropriate

EOPs,

or was adequately

resolved.

Secondly,

the attachments

and

EOPs were

compared to ensure that all

EOPs referenced

the appropriate

attachments,

and

that omissions

could not result in

a potential

over load of the

EDGs under

accident conditions.

An initial review of the

by the inspectors

noted that while many of the

JPE

recommendations

had

been

implemented

in

some

manner in the

EOPs,

there

appeared

to be

a number of omissions

and discrepancies.

An extended

meeting

was held

on Saturday,

Apri 1 5,

1986, with members of the

PUP to attempt to

resolve the identified concerns.

This was

a very constructive

meeting,

and

resulted in the resolution of several

of the concerns.

An example of an

concern

resolved

was the

requirement

in Section

2. 1. 1.2 of JPE-M-86-18 to

open breaker

MCCD-0825 prior, to resetting

a Safety Injection (SI) signal.

This step

was considered

necessary

to prevent automatic start of the

3S air

compressor

which could potentially exceed

the

EDG load limitations.

Since

all instrument air compressors

have

had the breakers

racked

out

and

danger

tagged to prevent starting, it was not necessary

to add this requirement to

the

During the

present

mode

of operations,

the

instrument air

compressors

have

been

replaced

by diesel air compressors

to prevent over-

loading the

In several

other

cases,

the

need to incorporate

a

JPE

recommendation

into

a specific

was resolved

because

the recommendation

was inserted into another

EOP which would be previously completed.

While the meeting

between

the inspectors

and

PUP personnel

resolved

a number

of EOP concerns,

other identified deficiencies

required additional revisions

of the

to in order

to

achieve

resolution.

These

deficiencies

are

considered

to

be significant in that they could have,

under

a given set of

circumstances,

resulted

in the

EDGs being overloaded.

Additionally, these

deficiencies

were

not detected

during the Plant Nuclear Safety

Committee

~'

(PNSC)

EOP reviews

and the procedures

were

approved for use

on

March 31,

1986.

Examples

of deficiencies

in the

approved

which could

have

permitted overloading

the

EDGs included the following:

JPE-M-86-18

required that for all

EDG scenarios,

the battery chargers

must

be aligned

and energized

no later than

30 minutes following a loss

of offsite power'.

This action

is essential

since

the

licensee

has

never

tested

the

capacity

of

the

safety-related

batteries

beyond

30 minutes.

The

JPE also requires

that the operating

pump

be

secured

within

20-30 minutes

of

the

event

and prior to

energizing

the

battery

chargers.

Allowing the

pump

to operate

beyond

30 minutes,

or

energizing

the battery

chargers

before

securing

the

pump,

could

overload the

3-EOP-E-O,

Reactor Trip or Safety Injection,

'approved

March 31,

1986,

Step

3.a

Response

Not

Obtained

(RNO)

directed

the

operator

to

3-EOP-ECA-0.0.,

Loss of'All AC Power.

ECA-0.0 was deficient in that it

did not ensure that

an

RHR pump would be secured

and

a battery charger

restored within 30 minutes.

ECA-O.O did not direct the operator

back

to

E-0

attachment

C which required energizing

the battery

charger s,

or attachment

D which required

securing

the

pump.

Attachment

E

of ECA-0.0 step 2.b contained di rection to energize

the battery charg-

ers,

but was inadequate

for several

reasons.

The entry condition for

step

2 was

an SI on Unit 4 and as

a result,

the direction to reenergize

a battery charger would not have

been

performed for an accident involv-

ing Unit 3.

Also, the directions

in step 2.b specified at 30 minutes

after reactor trip to verify adequate

diesel

capacity,

shed

loads

as

necessary,

then

place

a

battery

charger

in service

to

each

bus.

Starting at 30 minutes

and allowing

a reasonable

amount of time to shed

loads

per

attachment

D of E-0 would have placed the plant outside the

JCO

and

the

30 minutes of tested

battery capacity.

ECA-0.0 also did

not ensure

that

an

pump

was

secured

within 20 to

30 minutes

and

prior to reenergizing

a battery charger,

to prevent

overloading

the

Finally,

the

directions

for energizing

battery

chargers

in

ECA-0.0 Step 2.b did not specify utilizing the

"normal

bypass

switch"

for the

one

EDG available

case.

In this condition, utilizing the Reset

Switch, which is the normal

method of restoration,

could over load the

single

EDG by adding the boric acid tank heater

(15

KW) and boric acid

transfer

pumps

(27

KW).

In

response

to

these

identified deficiencies,

the

licensee

revised

ECA-O.O attachment

E on April 5,

1986.

Step

2 of attachment

E,

became

the directions for restoring battery chargers,

and

the directions for

manual

valving if an

SI occurs

on Unit 4 were

moved to step

3.

The

directions in step

2 were revised

to require that action

be

taken to

restore

the chargers

in

20 to 30 minutes,

and contain specific direc-

tions for reenergizing

the chargers

including the

use of the

bypass

~v

fl,

et

switch.

Although this revision significantly improved the procedure

and

reduced

the

chances

of overloading

an

EDG,

the fact that

the

battery chargers

must

be

energized

within

30 minutes,

and that

the

RHR pump must

be secured first, sti 1.1

needs to be emphasized.

JPE-M-86-18

Section

2. 1. 1. 10 requires that the

computer

room chiller

shall not be loaded onto

an

EDG until approximately

one hour following

'a loss of offsite power and after

a containment

spray

pump is secured

(55 minutes).

The delay in starting the computer

room chiller prevents

overloading

the

Licensee

test

data

indicates

a computer

room

rate of temperature

rise of 13'F/hr with no chiller on,

and controls

are

in place to maintain the normal

ambient temperature

at 67'F.

This

should

prevent

exceeding

80

F in the

computer

room

which is

the

temperature

above

which computer

card failures

have

been experienced.

3-EOP-E-0 dated April 31,

1986, did not ensure

that the

computer

room

chiller was

not started

for an

hour or unti 1 containment

spray

was

secured.

Attachment

E

step

2 directed

the

operator

to start

the

computer

room chiller within one hour of the reactor trip and loss of

offsite power.

Under these directions,

the operator could have started

the

computer

room chiller

(59

KW)

anytime

during

the . 60 minutes

following an accident,

potentially overloading

the diesel.

Attachment

D directions

were better.

They

required

the operator

to start

the

chiller in approximately

one hour,

and

noted that it should

be

done

after securing

the containment

spray

pump.

Due to the words "approxi-

mately" or "should",

however,

the potential for overloading the diesel

still exi sted.

In

response

to

the

identified deficiencies,

the

licensee

revised

3-EOP-E-0

attachment

D on April 5,

1986,

to prevent

the potential

of

overloading

the

Note

5 of attachment

D was revised

to require

placing the

B computer

room chiller in operation

in

55 minutes to

one

hour after

a reactor trip.

Attachment

E was not revised

and continued

to direct the operator

to start

computer

room chiller

B within

one

hour.

Since

step 3.c of E-0 directs

the operator to complete attach-

ment

E, the chiller could

be started

much earlier

than

one

hour

and

before

securing

the

containment

spray

pump.

Failure

to adequately

correct

this

deficiency

in all

will

remain

an

inspector

followup item (250,251/86-24-02)

pending resolution.

EOP-E-0

attachment

D is referenced

in several

EOPs for direction in

removing non-essential

loads

from the

EDGs during accident

conditions.

attachment

D of the version of E-0 approved

on March- 31,

1986, listed

46 separate

KW loads,

but did not differentiate

between

essential

and non-essential

loads.

An operator directed

to this attachment

t'o

remove non-essential

loads during accident

conditions

would probably

have

encountered

difficulty in determining

quickly which of the

46

listed loads were non-essential.

In addition,

several

essential

action

statements

such

as

securing

an

pump within 20-30 minutes

and

starting

a

computer

room chiller in

one

hour

were

contained

in

an

insert labeled "notes".

Placing actions

under "notes"

does

not appear

to

be consistent

with

NUREG 0899 or Westinghouse

Owners

Group guide-

lines for writing

Actions contained

under

"notes"

have

the

potential

to be overlooked during the stress

of emergency conditions.

In

response

to

the

identified deficiencies,

the

licensee

revised

attachment

D of E-0

on April 5,

1986.

The essential

and non-essential

loads

were

separated

which

should better

facilitate

the

removal

of

non-essential

loads during

an

emergency.

The action

statements

were

left in the insert

marked

"n'otes".

This item will remain

unresolved

(250,251/86-24-03)

pending determination

of the licensee's

commitment

to

NUREG 0899

and the Mestinghouse

PWR Owners

Group

EOP guidelines.

The inspectors

noted that the

EOPs contained

no directions to limit the

Unit 4 loads

added

to the

EDGs during

a

loss

of offsite

power

and

accident

on Unit 3.

The concern is that loads might be added to Unit 4

which could overload the'vailable

EDG, or reduce

the available

capacity

to support Unit 3 accident mitigation.

The licensee

revised

attachment

C

on April 5,

1986,

to

add

a caution:

"the unaffected

unit should

be aware or cautioned

on the effects of

EDG loading."

A Confirmation of Action letter

was issued to Florida Power and Light

confirming

EDG loading

commitments

ma'de during

a telephone

conferen'ce

on April 2,

1985.

Among the commitments

stated

were the following:

"Total loads

on emergency diesel

generators will be reduced to no

more than

2845 kilowatts per diesel

generator

and procedures will

be

changed,

and

operators

trained

on

these

changes

prior to

assuming duties, to assure

operation within this limitation."

A review of the

approved

on March 31,

1986,

indicated

inconsis-

tencies

in the

maximum

EDG load ratings stated.

3-EOP-ECA-0.2,

Loss of

All Power Recovery with SI Required,

contained

a caution that the loads

placed

on the energized

4

KV bus should not exceed

the capacity of the

power

source

EDG load rating of 2900

KW or 480 amps.

3-EOP-ECA-0. 1,

Loss of All AC Power Recovery Mithout SI Required,

contained

a similar

caution with an

EDG load rating of 2890

KW or 480 amps.

Both of these

EDG load ratings were in excess

of the 2845

KM listed in

the

Confirmation of Action letter

as

maximum

EDG loading.

Also,

a

maximum

EDG load test

conducted

in

May 1984,

only tested

the

EDGs to

2750

KW (reference

JPE-L-85-47).

The licensee

indicated to the inspectors

that the

2845

KW was

a maximum expected

EDG loading during an accident,

but that the 2950

KM for 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months

was the

FSAR emergency rating and was

addressed

in the

10 CFR 50.59 review.

The licensee

committed at

the exit following the first week of inspection to place

a maximum of

10"

2845

KW

EDG loading caution

in all applicable

This commitment

included

those

which

had

previously just

required

that

the

capacity of the power source

not be exceeded

with no value stated.

The

revisions

made to the applicable

on April 5,

1986,

did contain

cautions

to limit the

EDG loading to 2845

KW.

In addition,

however,

the caution

statements

also

allowed additional

loads to

be placed

on

the energized

4

KV buses until the red mark, or 168 hour0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months engine rating

(2950

KW) was reached

on the

KW meter for short periods.

This caution,

as inserted

in the

EOPs,

does not meet the load limitations in the

CAL

or the

verbal

commitment

made

to the

inspectors;

however, it does

fully meet

the

loading

limitations

specified

in JPE-L-86-59,

Revision

1.

This

apparent

failure

to

meet

a

commitment

to

the

Commission

was

due to the licensee'

failure to promptly notify the

Commission of a difference in understanding

of the commitment.

The

licensee

responded

quickly to resolve

most of the

EOP deficiencies

identified

during

the first week of inspection,

initiating

changes

on

Sunday,

April 6,

1986,

the

day of the exit.

These deficiencies,

however,

could have allowed operation outside of the

JCO and overloaded

the

EDGs with

a loss of all

AC power event.

This potential

should

have

been

detected

during tlie

PNSC

review

and

approval

process.

Indications

are

that

the

review, conducted

on the Saturday

evening just prior to the inspection,

was

directed

primarily towards

ensuring

each

of the

JPE

recommendations

was

incorporated

or resolved,

as

opposed

to ensuring

the

and attachments

could all interface

adequately

. to prevent

EDG overload.

The failure to

adequately

revise

EOPs to assure

that

EDG loading

was within limitations

pursuant to the Confirmation of Action letter issued

on April 2,

1986, is

a

deviation (250, 251/86-24-04).

7.

Training

The

inspectors

conducted

interviews

and

walkthrough

evaluations

of

licensed

operators

to

access

the

level

of training

on

the

EDG loading

restrictions

imposed

by JPE-L-86-59

and the revised

and attachments.

These evaluations

indicated that while the operators

were familiar with the

basic

contents

of the

attachments,

some

amount

of difficulty was

encountered

when they were walked through scenarios

requiring integrated

use

of the

and related attachments.

Part of thi s difficulty in utilizing

the

and attachments

was probably attributable to the deficiencies with

EOP and attachment

interface

as described

in Section

5.

Additionally, the

training

on these

revised

and

added

attachments

had

been primarily a

static

type of training.

Transparencies

were utilized to explain to the

operators

the

content

of each

attachment

such

as

re-energizing

battery

chargers

within

30 minutes

and

removing

non-essential

loads.

The'raining

on these

EOP revisions

and attachments

necessary

to support plant

operation

under JPE-L-86-59,

Revision

1

and JPE-N-86-18

was apparently

not

done

on

an integrated

basis

using

scenarios

on

a simulator

or plant walk-

throughs.

The

operators

expressed

the

opinion that earlier training

on

these

on the Standardized

Nuclear Utility Power Plant

(SNUPP)

simu-

lator had

been valuable,

but the

absence

of

a site specific simulator did

not allow this

type of training

on

the latest

revi sions.

Examples

of

specific deficiencies

in training included:

When placed

in certain

scenarios

such

as

a loss of all

AC power, the

operators

questioned

may not have performed

steps

necessary

to prevent

diesel

overload

such

as restoring battery

chargers

within 30 minutes,

or

removing

a

pump within

20-30 minutes.

This

was primarily

attributed to

a lack of references

between

the procedure

the operator

was

in,

and

the

attachments

to other

procedures

containing

these

actions.

The

absence

of integrated

training

on the

and attach-

ments also

appeared

to contribute to this problem.

The operators

were not adequately

familiar with the entry points into

EOP attachments.

Since

verbatim procedure

compliance

is required

and

the

lacked

appropriate

references

to attachments,

this lack of

familiarity had the potential

to place

the plant outside

the

analyzed

conditions in the JCO.

Some

steps

and sections

of the

EOP attachments

appeared

particularly

confusing

to the

operators.

An

example

was

the

action

statements

contained

within

an

insert

labeled

"notes"

in

attachment

D to

3-EOP-E-O,

Reactor Trip or Safety Injection.

The six "notes" contained

in this insert are designed

to be independent

and unrelated

actions to

be taken to prevent overloading

an

EDG in the

one

EDG available

case.

Due to the lack of integrated training

on the

use of these

attachments,

several

operators

thought these actions

were interrelated

and required

to be performed in the listed sequence.

Several

operators

demonstrated

a lack of familiarity with the locations

of controls necessary

to start the

computer

room chi llers

as directed

in note

4 of attachment

D to 3-EOP-E-O.

In isolated cases,

operators

indicated

a reluctance

to perform certain

actions

required

by

the

revised

and

attachments

which

were

contrary to the normal

mode of operation.

An example

was the require-

ment in step

3 of attachment

C to

3-EOP-E-0

to energize

a battery

charger utilizing the

"normal

bypass

switch".

Utilizing the

"normal

reset

switch",

which would

be

the

normal

method,

could result

in

overloading

the available

EDG due to auto

connected

loads.

At least

one operator indicated that

he

had been trained never to use the bypass

switch

and

hence

would

not.

This

reluctance

to

perform

actions

contained

in the

approved

and attachments

indicated

a

need for

operator

training

on

the

basis

for

unusual

actions

required

by the

w'Iky

l

12

The

licensee

promptly initiated

additional

training

on

the

upon

identification of the deficiencies

by the inspectors.

On-shift integrated

training in the

use of the revised

and attachments

was initiated

on

April 6,

1986.

A review the

second

week of the inspection

indicated that

the training

was

comprehensive,

and that it should resolve

the identified

concerns

in the area of EOP training.

The training included the use of the

attachments

and their'ntry points,

the

30 minute test limitations

on the

batteries,

the

basis

for

unusual

action

statements,

and

in-plant

walkthroughs of

E-0

and

ECA-0.0

under

simulated

loss of offsite power

with

an

SI

and loss of one

EDG or loss of all

AC conditions.

The large

number of questions

generated,

and the extended

length of several

of these

training sessions,

appear to indicate the additional training in use of the

revised

and

attachments

was

both

warranted

and

effective.

The

inspectors

verified that all on-shift licensed

personnel

had received this

additional

EOP training, but noted that the

licensed

personnel

attending

accelerated

requalification training

had

not received

the training.

The

licensee

committed

to train

these

licensed

individuals prior to their

assuming

licensed

duties.

This will be carried

as inspector followup item

(250, 251/86-24-05).

Training Brief

No.

122

was

written

on April 27,

1986,

to familiarize

licensed

operators

with

new

generation

and

the

incorporation

of

Westinghouse/JPE

recommendations

concerning

EDG loading.

Training

on

the

JPE

EDG loading

recommendations

was

based

on

Revision

0 of JPE-L-86-59

dated

March 29,

1986.

The inspectors

noted that Revision

1 of JPE-L-86-59,

dated

April 3,

1986,

contained

two significant

changes

from Revision

0

including:

Revision

0 of JPE-L-86-59

Section

3. 1.a required that

no

more

than

two

CCW or

ICW heat

exchangers

shall

be

in service.

Revision

1

allowed two or three

CCW or

ICW heater

exchangers

to be in service.

Revision

0 of JPE-L-86-59

Section

3. 1.d required that the

CCW flow to

components

in service. for the unit in cold shutdown shall-be throttled

until

pump

KW load is

less

than

291

KW or

a total

system

flow of

4600

gpm.

Section

3.2 required that

ICW flow shall

be throttled until

the

pump load is less

than

209

KW or a total

system

flow of 6000

gpm.

Revision

1 allowed changes

in the configuration of CCW and

ICW provided

that the total load for the operating

pumps

on the cold

shutdown unit

does not exceed

500

KW.

Gamma-Metrics Training

On July

15,

1985,

the

Commission

issued

an

"Order Modifying Licenses

to

Confirm Additional Licensee

Commitments

on

Emergency

Response

Capability"

confirming

the

licensee'

implementation

of -Regulatory

Guide

(RG)

1.97

modifications

in accordance

with the

schedule

commitments

finalized in- a

May 10,

1985 submittal.

The July 15,

1985,

Order confirmed that the

RG 1.97

modifications would be completed prior to the Unit 3 startup

following the

cycle

10 refueling

outage.

On

June

27,

1985,

the

licensee

submitted

a

letter to

NRR stating

that

the

Unit 3 installation

of the

neutron

flux

13

instrumentation

channels

(Gamma-Metrics)

was

complete

and the

system

was

operational

including

startup

testing,

plant

procedures,

and

operator

training.

A review of the

1985/1986 requalification curriculum and the

Gamma Metrics

neutron

flux -monitor requalification

lesson

plan during this

inspection

indicated that the licensee

completed this training for operators

licensed

prior to

February

1986,

as

part

of licensed

operator

requalification;

however,

a

review of the

replacement

Hot

License

Class

10

curriculum

revealed

that operators

receiving

licenses

after February

1986,

were

not

provided equivalent training

on the Gamma-Metrics

neutron monitor.

The

Gamma-Metrics

neutron

flux monitor

provides

reliable

neutron

flux

measurement

from reactor

shutdown to reactor full power level or from 10-'nv

to 10" nv in

a harsh

environment,

It is designed

to measure

neutron flux

with the detector

in

a high

gamma radiation

and electrical

noise

environ-

ment.

The

system

is

designed

to

operator

for

40 years

under

normal

conditions

and to survive

a design

basis

event

(DBE), providing reliable

measurement

before, during,

and after the

DBE.

On February

3-12,

1986,

the

NRC administered

reactor

operator

and 'senior

reactor operator

examinations

and requalifications

examinations

(Examination

Report

50-250/OL-86-01

and

Requalification

Examination

Report

50-250/

OL-86-01)

which contained

a

question

on

the

Gamma

Metrics

neutron

flux

monitor (RO 3.03,

SRO 6.05,

RO Requal

3.02,

and

SRO Requal 6.04).

Licensed

operators

admini stered

the requalification

examinations

were significantly

more

successful

on this question

than

the

Hot License

Class

10 candidates

reflecting the

Gamma Metrics instruction provided in 1985/1986 requalifica-

tion

training.

Subsequent

to

the written

examinations,

the

licensee

provided

the

following

comments

to

the

NRC

regarding

the

examination

question

on the

Gamma Metrics neutron flux monitor:

,"We request this question

be deleted for the following reason:

The

Gamma-Metrics

Monitor is

a

component

of the larger

safe

shutdown

system.

The Safe

Shutdown

system

has

not yet

been

turned

over for

plant

use.

A partial

turnover

of

the

Gamma-Metrics

Monitor

was

performed

in 1985.

The

Gamma-Metrics

Monitor was

functional

at

the

time but was to

be

used for indication only.

Procedure

changes

and

training were determined

to not

be necessary.

Full training will be

implemented

at

time, of safe

shutdown

system

turnover.

A training

overview was presented

to licensed operators

as "look ahead"

to inform

them

of the

new

instrument

in their control

room.

However,

the

instruction

given

was

not detailed

because

of the limited purpose

of

the Gamma-Metrics Monitor at that time."

Based

on the licensee's

response

to,the

Gamma-Metrics

neutron flux monitor

examination

question,

the

NRC

deleted

this

question

from the

reactor

operator

and

senior

reactor

operator

examination

and

requalification

examination.

The deletion of this question affected grading statistics

and

impacted the

NRC's granting of licenses

and renewal

licenses.

,I

r ly

S

b

The licensee

indicated that Hot License

Class

10 candidates

were not trained

on the

Gamma-Metrics

system

because

of the training staff's

misconception

that

the

Gamma-Metrics

system

was

not fully operational

and therefore

training

was

not

required

to

be

completed.

This

misconception

was

apparently

due to the training staff's

conclusion

that the portion of the

Gamma-Metrics modifications installing

a neutron flux channel

to the alternate

shutdown

panel

was

incomplete.

The

Gamma-Metrics

neutron

flux channels

installed in the control

room pursuant to the July 15,

1985 "Order Modifying

Licenses

to Confirm Additional Licensee

Commitments

on

Emergency

Response

Capability" were operational.

During a previous inspection

conducted

by the resident

inspection staff, the

inspectors identified that prior to August 14,

1985,

no operator training on

the

use of the

new nuclear

instrument

channels

had

been

performed.

Only

after

NRC inspectors

informed the licensee

that their June

27,

1985 letter

stated

that training

had

already

been

completed,

was

a requalification

training lesson

plan developed

and implemented.

In that

the

Gamma-Metrics

neutron

flux monitor

was

operational

in the

control

room,

required

in the

emergency

procedures

(4-EOP-FR-S.2,

Response

to Loss of Core

Shutdown,

Step 2),

and provided the only reliable

neutron

flux measurement

in

a harsh

environment pursuant to Regulatory

Guide 1.97,

the inspectors

consider that training was mandatory for Hot License Class

10

candidates.

10 CFR 50,

Appendix B, Criterion XVI requires

that in the case of signifi-

cant conditions

adverse

to quality, that

measures

shall

assure

that

the

cause of the condition is determined

and corrective action taken to preclude

repetition.

Contrary to the

above,

the licensee's

comments

on tlie Gamma-

Metrics neutron flux monitor examination

question

and the licensee's

failure

to provide

Gamma-Metrics training to Hot License

Class

10 candidates

were

indicative of

a

general

failure to take

adequate

corrective

actions

to

preclude repetition of deficiencies identified during

a previous

NRC inspec-

tion.

This is

a violation (250, 251/86-24-06).

In response

to this inspection

finding, the licensee

committed to provide

training

on

Gamma Metrics to those

Hot License Class

10 candidates

who were

successful

on their

NRC license

examinations

and who currently hold reactor

operator

and senior reactor operator licenses.

Implementation of

EDG Loading Compensatory

Controls

To comply with the Justification

for Continued

Operation with One Unit at

Power

and

One Unit in Cold Shutdown Relating to Emergency

Diesel Generator

Loads (JPE-L-86-59,

Revision 1),

and (JPE-M-86-18,

Revision 0), the licensee

established

clearance

number 86-3-166.

This clearance

allowed the placement

of danger

tags to ensure

the proper position for components

necessary

to

prevent

overloading

the

The

inspectors

verified

through

plant

walk-downs that all components listed in the

JCO were correctly positioned

with the required administrative

controls

(tags)

in place.

The inspectors

also verified the following JCO recommendations:

i '

15

Section 3.0 required that the total

pumping loads

on Unit 4 CCW/ICW be

limited to

500

KW to provide additional

EDG margin.

Section

3. 1

and

3.2

provided

a list of

CCW

and

ICW valves

supplying

non-essential

Unit 4

loads

that

should

be

closed

to

ensure

that

the

500

KW pump

loading

is

not

exceeded.

The

inspector

verified

closed

and

administratively controlled all Unit 4 valves listed in Section

3. 1 and

3.2.

A danger

tag,associated

with clearance

86-3-166

was

found with

the

wrong clearance

number

and

was promptly resolved

by the licensee.

Additionally, the inspectors

were concerned that the level of approval

to remove the

EDG loading clearance

was not sufficiently high enough in

the

licensee's

management

organization.

Subsequently,

the

licensee

added

the Operations

Supervisor

and the Operations

Superintendent

to

the clearance

for approval.

This resolved

the inspector's

concerns.

Sections

3. l.d and 3.2.d required that the

CCW/ICW flows to the Unit 4

components

in service

be throttled unti 1

the

combined

pump

loads

do

not

exceed

500

KW.

The

inspector

verified that

the

ICW/CCW heat

exchanger

valves

were throttled

to

ensure

that

the

500

KW was

not

exceeded

and were administratively cont'rolled.

The inspector indicated

to the licensee

that locks might provide

more positive control

than

danger

tags over these throttled valves

and

the licensee

committed to

install locks.

JPE-M-86-18,

Revision

0,

Section

2. 1. 1. 1, required that

each control

room

KW meter

and

ammeter

be

marked to help ensure

that the

load limits were

not exceeded.

The inspector

verified that the con-

trol room meters

were

marked

in accordance

with the table

in Section

2. 1

~ 1. 1.

The

maximum

KW and

ampere

ratings were marked with red tape

and the normal

maximum loading of 2845

KW was marked with orange

tape.

It should

be noted,

however, that the red marker

KW values

ranging from

2910

KW to 2960

KW were in excess

of the

maximum value of 2845

con-

tained in the

CAL.

JPE-M-86-18,

Revision 0,

Item 2,

requires

that

the

instrument

air

compressors

be

deenergized

to prevent

overloading

the

and

be

replaced

by portable

diesel

compressors.

The inspector verified that

the electrical

breakers

for all instrument air compressors

were

racked

out and tagged

under

a clearance.

JPE-M-86-18 Revision,

0 Item 2. 1.3. 1, requires that the Unit 4 high head

safety injection

(HHSI) and containment

spray

pumps

be prevented

from

starting to prevent overloading the

The inspector verified that

these

pumps were in the pull-to-lock position

and caution tagged.

JPE-L-86-59,

Revision

1, Item 2.2, requires that the computer

room tem-

peratures

be limited to 67

F during

normal

operations.

At a

13 /hour

rate of temperature

rise, this limitation will allow a one. hour delay

in restarting

.the

computer

room chiller on

a

loss

of offsite

power

reducing

the initial EDG loading requirements.

The inspector verified

the

computer

room temperature

to

be

less

than

67'F

and

that

the

ql '

16

temperatures

are

logged

twice

per

shi ft to

ensure

thi s limit i s

maintained.

JPE-M-86-18,

Revision

0,

Item

2. 1. 1. 1

recommends

that

the

KW and

ammeters

for the

be calibrated

on

a monthly basis to ensure

the

are

not

overloaded.

The

inspector

reviewed

the

initial

calibration

pack'ages

for the

A and

B

(Work Orders

058980

and

047797 .respectively)

that were

completed

on

March 24,

1986

and

they

appeared

acceptable.

In addition,

the monthly calibration tests

were

under development

and scheduled

to be

implemented within one

month of

" the initial calibrations.

A review of the monthly calibrations will be

an inspector

followup item (250, 251/86-24-07).

The licensee

determined

that

on

a loss of the

"B" EDG, that

a single

failure of battery

3B could prevent

the automatic transfer of motor

control center

(MCC) "D" to the

"A" EDG.

MCC "D" supplies

essential

loads

including

two battery

charger s

and

an

emergency

containment

cooler.

A modification (PC/M-86-047)

was completed to allow a transfer

of

MCC

"D"

on

a loss'f

the

"B"

even without battery

"3B"

available.

The inspector

reviewed this modification package

and the

corresponding

safety evaluation

and they

appeared

adequate

to resolve

the identified deficiency.

On April 14,

1986 at

1:05 a.m.,

the licensee

deenergized

the

4B 4160 volt

bus to perform Appendix

R upgrade modifications

on Unit 4.

The deenergized

4B 480 volt load center,

which is normally powered

by the

4B 4160 volt bus

was reenergized

from the

4A 4160 volt bus via crosstie

breakers

with the

4A

480 volt load center.

At 5:20 p.m.,

approximately

16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months s after entering

this

abnormal

alignment,

the

licensee

restored

the

normal

power

supply

alignment to the 480 volt load centers.

Supplying

power

to

the

4B

480 volt

loads

via

the

4A

4160 volt bus

represented

an

emergency

diesel

generator

(EDG)

loading

case

that

was

unanalyzed

in the licensee'

Justification for Continued Operation with One

Unit at

Power

and

One Unit in Cold Shutdown

Relating to Emergency

Diesel

Generator

Loads (JPE-L-86-59,

Revision

1) transmitted

to the

NRC by letter

dated April 8,

1986.

In that safety evaluation,

the licensee

assumed that

the

4160 volt busses,

480 volt load

centers

and

480 volt motor control

centers

remained

in their normal configuration

as described

in

FSAR Section

8.2.

Energizing

the

4B loads

via

the

4A 4160 volt bus violated

these

assumptions

and

placed

unanalyzed

loads

on the

4A bus which potentially

could have

exceeded

loading limitations placed

on the

"A"

EDG specified

in

JPE-L-86-59,

Revision

1.

A subsequent

safety evaluation

(JPE-L-86-63) of the April 14,

1986 alignment

of the

4B

loads

and their effect

on

the total

EDG loading

assumed

in

JPE-L-86-59,

Revision

1 concluded that the post-accident

loading of the "A"

EDG would not have

exceeded

the loading

assumed

in JPE-L-86-59,

Revision

1.

JPE-L-86-63

identified the

"C" boric acid

tank

heater

and

the

hydrogen

analyzer

heat tracing

as the only loads which would have

auto

connected

to

the

"A" EDG.

This represents

an additional

21

KW to the

"A" EDG loading;

17

however,

during the time that the crosstie

was in effect,

the

4D normal

containment

cooler,

also

powered

from the

4B load center

and representing

64

KW, was maintained deenergized

with breaker clearances

in place.

There-

fore,

the net result of this specific

EDG loading

case of April 14,

1986,

represented

an

actual

43

KW reduction

below

the

loading

assumed

in

JPE-L-86-59,

Revision

1.

The safety evaluation

conducted

by the licensee

pursuant

to

10 CFR 50.59

provided justification for continued operation of Unit 3 with Unit 4 in cold

shutdown relating to

EDG loads

and concluded that by following the actions

and recommendations

specified in JPE-L-86-59,

Revision

1, that post-accident

loading

would

be

limited to

2845 kilowatts

per diesel

generator.

Contrary to

10 CFR 50.59, for approximately

16

hours

on April 14,

1986,

the

licensee

added

additional

auto

connected

loads

to the

4A

EOG beyond

those

analyzed

in JPE-L-86-59,

Revision

1, without first evaluating their

impact

on

EDG loading.

This is

a violation (250, 251/86-24-08).

The crosstie

of the

4B 480 volt load center to the

4A 480 volt load center

was

accomplished

under

clearance

number

86-4-063

and

in

accordance

with

approved operating

procedure

4-0P-006,

480 Volt Switchgear

System to support

PCM 83-155 (Appendix

R Modifications).

Operating

procedure

4-OP-006 did not

incorporate

necessary

procedural

instructions,

nor

did

the

licensee

established

administrative

controls (i.e.,

tagging

the tie breakers)

to

preclude

invalidating the safety evaluation

in JPE-L-86-59,

Revision

1 by

cross-tieing

480 volt load centers.

In addition,

the

licensee

had

not

sufficiently sensitized

shift supervisors

to the

requirement

for

a formal

evaluation

prior to

adding

loads

to the