ML18004B895

"D. Tibbitts, Supervisor,

Regulatory

Compliance

B. Tomlin, Senior Engineer,

TS

~M. Wallace, Specialist,

Regulatory

Compliance

~E.

Willet, Manager, Modifications (HPCS)

Other

licensee

employees

contacted

included

construction

craftsmen,

engineers,

technicians,

operators,

mechanics,

and office personnel.

NRC Resident

Inspectors

"George Maxwell, Senior Resident

Inspector

~Attended exit interview

Exit Interview

The inspection

scope

and findings were

summarized

on June

5,

1987, with

those

persons

indicated in paragraph

1 above.

The inspector

described

the

areas

inspected

and

discussed

in detail

the

inspection

findings.

No

dissenting

comments

were received

from the licensee.

Violation,

Failure

to

perform

adequate

safety

evaluations,

para-

graph 8.

The licensee

did not identify as proprietary

any of the materials

provided

to or reviewed

by the inspector during this inspection.

3.

Licensee

Action on Previous

Enforcement Matters

This subject

was not addressed

in the inspection.

4.

Unresolved

Items

Unresolved

items were not identified during this inspection.

5.

Licensee Action on Previously Identified Inspection

Findings (92701)

,a.

(Closed)

Inspector

Follow-up Item 400/86-53-08:

Surveillance

Program

Procedures

This

item identified

an

incomplete

surveillance

program

in that

not

all

procedures

required

to

perform

Technical

Specification

surveillance

requirements

had

completed

the

review

and

approval

process.

The

inspector

verified

those

specific

procedure

types

identified

in

the

initiating

report

were

presently

completed.

Additionally the

inspector

performed

a

sample

check

to verify

technical

specification

surveillances

were

scheduled

on

the

master

schedule

and that procedures

were approved to perform the tasks.

b.

(Closed)

Inspector

Follow-up Item 400/86-53-12:

Identification

and

Control of Commitment Related

Process

Instrumentation

This

item identified

an

incomplete

program for installed

process

instrumentation

used

for safety-related

or technical

specification

functions in that not all applicable

instrumentation

was identified

or entered

into

a calibration program.

The licensee

has developed

a

Required

Instrumentation

List (RIL) to identify instrumentation

required

to

meet

Final

Safety Analysis

Report

(FSAR) or technical

specification

requirements.

The list used

in conjunction with the

technical

specificati.on

Cross

Reference

List

(PGO-031)

provides

traceabi lity from technical

specification

survei llances

to specific

instruments

and the calibration procedures

for the instruments.

The

licensee

stated that the out - of - tolerance

instruments of the RIL

would be evaluated for previous

useage

per the individual instrument

cal,ibration procedures.

The inspector

reviewed the completed instrument

list and performed

random checks to verify process

instruments

were

included

as required.

Discrepancies

were not identified in this sample.

(Closed)

Inspector

Follow-up Item 400/86-33-06:

Onsite

Packaging of

Equipment

and Materials

This item identified the

need for the licensee

to direct attention to

packaging of spare parts

and material to prevent

damage

during trans-

ient from receipt

inspection

and

warehouse

storage.

The inspector

reviewed

the following procedures

to verify the programmatic guide-

lines for onsite packaging:

HNP-9.3.1,

Responsibilities

for

Packaging

Shipping

Receiving

Handling,

Storage

and

Maintenance

of Spare

Parts,

Material

and

Components

Mhile

Under

the

Control

of

Marehouse

Personnel,

Revision 0.

TMM-110, 'Receiving,

Packaging,

Storage

and Maintenance

Require-

ments for Spare

Parts Material

and Components,

Revision 0.

These

procedures

appeared

to supply adequate

guidance for protection

of

spare

parts

and material

onsite.

In addition,

the

inspector

'discussed

the potential

for damage

of parts

and

materials

with

warehouse

personnel.

Problems

were

not identified in this

area

and

i

appeared

that the guidelines

of the

above

procedures

were

adequately

implemented.

6.

Field Change

Requests

As

a result of the (}uality Assurance

preoperational

review of the licensee

Design Control

Program,

areas

were identified which required

a review to

a greater

depth,

especially with respect

to implementation.

These

areas

included Field

Change

Requests

(FCR)

and

Plant

Change

Requests

(PCR).

Field

Change

Requests

were

changes

to plant design

processed

during the

const, uction

phase utilizing guidelines

of the construction

gA program.

Following the establishment

of

a

system

as

"operable"

any

changes

were

required to

be

made via the

PCR process.

This process

required

addi-

tional

reviews

of the

specific

change

for maintenance,

operations

and

training impact.

During inspection

50-400/87-09

the inspector identified

a number of modifications being processed

via the

FCR process

although the

associated

systems

had been declared

operable

and released

to operations.

This inspection,

in part,

reviewed this

aspect

of the

design

control

program with regards

to volume of FCRs

open

or processed

following issuance

of the

low power operating

license

and

the actions

of the licensee

to

evaluate

and closeout

remaining

FCRs.

The

FCR procedure

was cancelled

on

April 21,

1987,

which ended

the mechanism for initiating future

FCRs.

The

licensee

identified 829

FCRs which were

open or resolved following issuance

of the

low power operating

license

on October 24,

1986.

Of these,

606

were resolved prior to the specific system's

operational

date

and included

in the system

release

to operations.

These

FCRs appeared

to be appropri-

ately

processed.

Systems

under

construction,

i.e.,

waste

treatment

systems,

accounted for 25

FCRs which were properly addressed

as

FCRs.

The

remaining

196

FCRs received

a multi-discipline review by Technical

Support,

associated

discipline,

and

Environmental

and Radiological

Control

(E&RC)

to determine

impact

on plant systems,

procedures,

operations,

and training.

The result of this review was the identification of ten

FCRs with potential

impact

on

system

operations,

tests,

or training.

Of this ten

FCRs,

two

were identified

as safety-related.

The safety-related

FCRs

and

impact

were

as follows:

1)

FCR-I-1735

impacted

on training notes

and was

a document

change

only.

2)

FCR-I-3952

impacted operations

procedures

with respect to

a set

poin

change.

The following is

a list of the eight non-safety

and their

impact:

FCR-I-3248,

Revision 2,

impacted

a

system description

which is due

for issuance

in August 1987.

FCR-SI-847,

Revision 2,

required

adding

computer points to scaling

documents

SCN 034 and

SCN 036.

FCR-HV-1728 and

FCR HV-1730 involved minor installation details which

impacted information in training notes.

FCR-I-3962

and FCR-I-400 impacted setpoints

and

a procedure revision.

FCR-I-3995

involved

an annunciator

window modification which changed

a procedure

reference.

FCR-l~i-2032

affected

the

chemical

feed

system

and

impacted

class

training notes.

Additionally, 101 of the

196

FCRs

open after

system turnover remain open;

36 of these

were partially implemented.

Those

FCRs not under

implementa-

tion will be deleted,

converted

to PCRs,

or receive

an equivalent review

and evaluation

process

similar to the

The remaining

FCRs will be

incorporated

into

an

implementation

schedule

by July 15,

1987.

The

following activities

by

the

licensee

demonstrate

positive

aggressive

action to resolve

the

FCR issue:

1)

Cancellation

of the

FCR procedure

to prevent further initiation

of FCRs.

-2)

Identification of the

scope of this issue

through review of FCRs

and associated

system

impact.

3)

Organization

and completion of multi-discipline reviews of each

open

FCR.

For the

sample

reviewed,

the licensee

appeared

to have achieved

adequate

control of the

FCR process

to provide

a timely and sufficient resolution

of potential

problems

and closeout of the

FCR program.

Per

CP8L (Harris

Nuclear .Project) letter

NRC-555 dated

May 1,

1987, (File

Number SHF/10-

10000),

the licensee

had committed to review the entire

FCR process.

7.

Plant

Change

Requests

(PCRs)

-- Implementation

The

inspector

reviewed

selected

to determine if adequate

design

controls

were

utilized to

ensure

appropriate

reviews

were

performed,

drawings

were

accurately

revised,

safety

evaluations

were

correct

and

thorough,

required

training

was

complete

and timely,

and modifications

were accurately installed.

The following PCRs were reviewed:

PCR 0

REVIEWED

TITLE

81

103

187

234

253

259

292

301

353

423

466

467

489

667

668

785

824

825

842

854

932

1026

1119

1145

1157

1190

1191

1286

1388

1515

1571

7.5

KVA Instrument

Power Supply Inverter

Demineralizer

Bypass

Around

CST and

R'MST

I/P Transducers

Indicator Scale

Changes

Appendix

R Power

Panel

Protection

Auxiliary Feedwater

(AFM) Venting Provisions

Design

Change - Turbine Driven AFM Pump Oil

7.5

KVA Inverter Ferro-Resonant

Transformer

Valves Not on Plant Drawings

RHR/SI System

Vent Valves

AFM Pump

Low-Low Suction Pressure

Switch

AFM Pump Suction

from CST

NRC Concern

ESM/AF

Containment

Spray Niniflow Orifice Replace-

ment with Throttle Valves

AFW Pump Vents

Service Air Leakage

Outside Diesel

Generator

Building

AFM Pump

Th. ust Bearing Coil Spring

Evaluation of

PCR 824

AFW Pump Bearing Walkdown

RMST Transmitter

Replacement

Essential

Chiller Recirculation

Seal

Piping

AH(1A-SA) E-28(1A-SA)

Thermolag Interference

AFW Check Valve

Condensate

Storage

Tank

AFW Pump

Low Flow Trip Delay

Instrument

Loop F-113 Controller Card Update

AFW Check Valve Backleakage

Problem

AFW Check Valve Backleakage

Steam

Hammer Line number

3BD4-13-SN-1

Steam Generator

Blowdown Valve-Temporary

Repair

The inspectors

found no evidence that

PCRs were not receiving appropriate

reviews or that single line electrical

and flow diagrams did not accurately

depict the as-built configuration of the plant. It was noted that drawing

details

are not always revised in a timely manner which could lead to work

being performed without updated

drawings;

however,

the inspectors

did not

identify any evidence

of this.

The

safety

evaluations

and

evaluations

included

in the

selected

PCR packages

were reviewed

by the inspectors.

It appeared

that

the justification for the conclusions

reached

in the evaluations

were not

always well documented.

This is addressed

in more detail

in paragraph

8

of this report.

The inspectors

reviewed the training provided for modifications installed

under

the

and

FCR programs.

The training of the onshift operators

appeared

to

be

adequate;

however,

there

appeared

to

be

an

excessive

amoun

of

time

required

to

include

modifications

.in

the

operator,

qualification

and requalification training.

Specifically, the inspectors

reviewed

the training

guide

and

system

description

for

the auxiliary

feedwater

system;

neither

of

which

completely

addressed

the

recent

modifications to the system.

The inspectors

field verified

system modifications

and drawing accuracy

by extensive

walkdowns of selected

The following PCRs

were field

verified.

FIELD VERIFIED PCRs

153

Demineralizer

Bypass

Around

CST and

RWST

259

AFW Feedwater

Venting Provisions

292

Design

Change

Turbine Driven AFW Pump Oil

423

RHR/SI System

Vent Valves

466

AFW Pump

Low-Low Suction Pressure

Switch

467

AFW Suction

From

CST

489

tiRC Conce. n

ESW/AF

667

Containment

Spray MiniflowOrifice Replacement

668

AFW Pump Vents

785

Service Air Leakage Outside Diesel Generator Building

932

Essential

Chiller Recirculation

Seal

Piping

1145

AFW Check Valve Accessibility

1190

Feedwater

Pump

Low Flow Trip Delay

1286

AFW Check Valve Backleakage

Problem

1388

AFW Check Valve Backleakage

Temporary Modification

1571

Steam Generator

Blowdown Valve Temporary Repair

There

was,

no evidence

of any discrepancies

between

the as-built config-

uration'nd

the modification.

PCR Package

Contents

The inspector

examined

the methodology,

interfaces,

and output associated

with work package

development for mechanical

After a design

package

has

been

developed

by

engineering

and

administratively

processed

by

Technical

Support, Harris -Plant Construction

Section

(HPCS) developed

work

packages

for implementation

of the modification.

The

packages

include

work instructions,

additional

drawings, guality Control

hold points,

and

material for the job.

The inspector

examined

the following new procedures

that had

been written

for package

development:

0'

PROCEDURE

MAP-1

TITLE

Harri s Pl ant

Modificati on

Unit of Conduct of Operations

REVISION

MAP-2

MAP-3

MAP-4

Preparation,

Approval,

and

Control of Modification Unit

Procedures

Training of Harris Modification

Unit Personnel

Process

Control

0

These

procedures

were issued

in May 1987

and were in the process of being

implemented.

Site

personnel

stated that these

documents

varied minimally

from previous construction

procedures.

These

procedures

reference

some

existing

pre-operational

procedures.

was

phasing

out

the older

referenced

procedures

and appeared

to have firm control of that evolution

appeared

to

interface

well with the

other

plant

groups.

personnel directly interfaced with trades

personnel

in work accomplishment

and,

thus,

were

spokesmen

to the other plant groups

on job status.

When

required,

routed

changes

to modifications

and to

HPES.

was

developing

a working relationship with the Technical

Support group.

The

inspector

reviewed

several

modification

work

packages

with

personnel.

With the

personnel

involved,

the

process

appeared

to

be

thorough.

With so few PCRs in process

and with the

new procedures,

was

taking their

time with the first few packages

in order

that the

mechanics

of the package

creation could be understood.

The

inspectors

examined

the

following

packages

for content

and

completeness:

NO.

TITLE

1134

1145

Flow Element

Flowing from

Duct Work Lining

AFW Check Valve Accessibility

310

Vendor Mobile Solidification

System

Hookup

These yet-to-be

closed

packages

appeared

to

be

complete

to the point of

the exceptions

noted in the packages.

Some of the Field Revisions in

1145 were difficult to follow, but all packages

had the essential

elements

required

by site

and regulatory documents.

The inspectors

extensively

reviewed

1286, Auxiliary Feedwater

(AFW)

system

modi> ication to prevent back-fill in auxiliary feed lines,

and

842,

AFW

pump thrust

bearing

coil

springs.

The.

remaining

paragraphs

discuss

these

1286 installed

three additional

check valves in the

AFW system motor

driven

pump train

(one

per line to each of the

steam

generators).

The

check valves

were to prevent

back

leakage

from the

steam

generators

and

main

feed lines..

The

two previously existing

check valves

in each line

(six total) did not prevent

the

back

leakage.

The primary check valves

(Pacific, Inc.) in each line which were adjacent

to the

steam

generators

that

were

to prevent

leakage

had

been

previously unsuccessfully

worked

(PCR

1145)'.

Each of three

new valves

(Anchor-Darling,

Co.)

were welded

into the

system

between

the two existing check valves in each line.

In

discussions

with site

personnel,

. the

following options

for

new

Anchor-Darling valves

had

been

considered:

replacement

of the Pacific

valves

nearest

the generator

with the new-valves,

using 'flanged

valves

instead

of welding

the

valves,

using

hangers

on

the

proposed

flanged

valves,

using

another

valve type,

using soft seated

valves,

leaving the

existing Pacific valves in place in case

soft seats

were eventually

used

(to prevent

thermal

deterioration

of the soft seat),

and installing

a

motor operated

valve in lieu of

a check valve.

Due to operational

time

limits, the welded. valve configuration

was chosen.

None of the above

two

paragraphs

of background

information was present

in the

PCR package.

The licensee

had contingency

plans

should

the

new Anchor-Darling valves

fail to prevent

back leakage.

As indicated

above, soft seats

for the

new

Anchor-Darling valves

and motor operated

valves were considered.

The site

had ordered

soft seat kits (a

new

PCR would

be required)

for the

newly

installed valves

but,

per the licensee,

they

had not ordered

hard disks

for the existing valves.

The

new valves

appeared

to

be preventing

back

leakage.

Aside from the

operability/functional tests for the system,

the

new valves

had

not,

been

challenged

with

system

transients.

Per

the

licensee,

operational

procedures

had

been

changed

to prevent the

usage of the

AFW system during

startup

which

would limit the

cycling

of

the

system

check

valves.

Further,

AFW

system

testing

(surveillance)

and

usage. will determine

whether or not the modification will be viable.

Due to the

awareness

of historical

operation

problems with check'valves

'nd

the likelihood that

the

new

Anchor-Darling

valves will probably

require cyclical

preventative

maintenance

(PM), the inspector

discussed

valve

PM with the licensee.

NRC Information Notices

such

as

86-01

and

86-09

had

addressed

problems

with check

valves

in feedwater

and

AFW

systems.

It has

been historically recognized

that

check

valves

leak,

internal

parts

separate

from the valve body,

and the valves

have active

fai lures (fail to open).

Per discussion with the Technical

Support

Group

AFW system engineer,

he was

unaware of any

PMs on the valves,

but directed

0

the inspector

to the Maintenance

Group.

The Maintenance

Manager

was aware

of

an

Ins itute

of

Nuclear

Power

Operations

Significant

Operating

Experience

Report

(SOER 86-3)

which Harris

Plant

Engineering

Section

(HPES)

was

reported

to

be addressing.

The maintenance

manager

said that

no

PMs

were

currently

being

performed

on

check

valves.

Per

HPES

personnel,

they were in the preliminary stages

of developing

background

information towards implementation of PMs.

and Section 6.5. 1.4. 1 of the Plant's Technical Specification

(TS) require

that

a

safety

evaluation

be written

should

a

change

or

modification to

a

safety

system

occur.

Both

documents

require

that

certain

questions

be

addressed

in preparing

the evaluation.

One of the

questions

is whether the change

or modification represents

an increase

in

probability

of

occurrence

or

the

consequences

of

an

accident

or

malfunction.

and

the

TS state

that

the

evaluation

shall

include

a written determination,

with bases.

Procedure" AP-011,

Safety

Reviews,

Revision

1,

implements

the

TS

and

requirements.

Paragraph

5.2.c,

of the

procedure

addresses

what

the

evaluator

shall

analyze

addressing

the

above

question.

The

paragraph

states

that

an

action which makes

even

a single failure to safety related

equipment

more

probable

must

be considered.

1286,

(AR( check

valves)

did contain

a safety evaluation

(Form 2,

pages

15 to 17).

Part 2, paragraph

7.3, of the evaluation

which restates

the probability increase

question

of AP-011

was

answered

"no"

and

the

bases

for this

answer

was

stated

to

be that

the modification to the

auxiliary feed

system will not affect the design

conditions of safety-

Answers

to

other

questions

in evaluation

were

similarly terse.

310 and

489 had extensive

safety evaluations

which

more formally addressed

the design

envelope of the modification that they

performed.

Neither the

1286 evaluation

nor the

package itself

contained

any information or analysis

that di scussed

this extra internal

check valve parts (disk, disks pin,

and fasteners)

and/or

valve failure

potential

caused

by the valves'ddition to the

AFM system.

Aside from the

one line answer to paragraph

7.3, Part 2, of the safety evaluation,

there

was

no information from which to draw the conclusion that the three check

valves installation did not increase

the probability of system failure or

malfunction.

The above identified inadequate

safety evaluation is but one example.

825 contains

another

example of an

inadequate

evaluation,

the specifics

of which

are

detailed

below.

Both

examples

represent

a violation,

400/87-20-01,

of 10 CFR 50.59,

Inadequacy

of Safety Evaluations.

There

were three

similar

generated

by the site

on the

AFM pumps.

They are

as follows:

0

10

ISSUED

DATE

NUMBER

FORM

1

INITIATION

SAFETY

EVALUATION DATE

PUMP AFFECTED/

COMMENTS

824

1-17-87

and

4-27-87

1A - SA/

Temporary

Modification

825

1-17-87

1B-SB and

1X-SAB/

Evaluation

842

1-20-87

2-3-87

All Three

Two modifications

(PCR

824

and

842)

were to

remove three thrust springs

from

each

of the three

AFN pumps

and

one modification

(PCR825)

was to

evaluate

not performing the spring removal unti 1

a later date.

The

pumps

were modified on dates

as follows; lA-SA on 1-17-87,

1Ã-SAB on 2-5-87,

and

1B-SB

on 3-15-87.

Per the licensee,

842

was to provide closure

of

work on all the

pumps

but maintenance

had yet to forward the

completed

work packages

to Technical

Support

Group (not in the

PCR package).

824

was

to

perform

a

temporary

modification

on

the

1A-SA

pump;

a

Nonconformance

Report

(87-012)

which was still open at the

time of the

inspection

indicated that the work had been

performed prior to issuance

of

the

PCR and work ticket.

The initiating impetus

to all

pump work was the fact that the

1A-SA pump

did

not

meet its

pump

curve

during operability testing.

There

were

several

contacts

made with the

pump vendor

( Ingersoll-Rand),

some of which

were not documented

in the

PCR packages.

PCR 825 references

a

memo from

the vendor

and

telephone

conversation

with. the

vendor

and documentation

for these

were not in the

PCR packages.

PCR 842 contained instructions to

remove 'the thrust springs

from the

pumps

and replace

them with a spacer.

825

which is

an

engineering

evaluation

with

a

safety

evaluation

included

provided justification for continued

operation

of

the

two

unmodified

pumps.

The engineering

evaluation

states

that the insufficient

head

problem (lA-SA pump)

was

analyzed

to

be attributable

primarily to

improper assembly

and/or

missing parts

and secondarily to abnormal

wear of

the balancing

drum.

The analysis

was not

a par t of the

PCR package

unless

it

was

the

referenced

vendor

memos

The

discovery

findings

at

the

disassembly

of the

1A-SA was not .available

in the three

PCR packages.

The safety

evaluation

( Form 2,

pages

15 to

17)

found in

PCR 825 did not

adequately

support continued

operation

of the

two unmodified

pumps.

The

Part 2, paragraph

7.3 question of the safety evaluation

as4ed

the question

as follows; Mill the probability of occurrence

of malfunction of equipment

important to safety

be increased?

0'

11

The utility answered

the question

as

no

and

the

bases

was that

AF'M pump

and

system reliability is not affected.

Although the two unmodified

pumps

were

not exhibiting

low head output

as the

pump

1A-SA,

no explanation

as

to the potential

missing

or rubbing parts

in the

unmodified

pumps

was

given.

The removal

and replacement

of the

1A-SA springs

was thought to be

a reduction

in probabi li y of

a malfunction

(and

a repair

method).

As

with