ML18005A257

1874

1515

PCR 2335

PCR 2298

PCR 2046

1286

1391

1837

PCR 947

PCR 913

1587

PCR 0174

PCR 2271

1435

PCR 2292

PCR 2005

PCR 2286

Damaged

Support

Steam

Hammer

Damage

Valve

1BD 30 Closure

Time

Blowdown Valve Cracked

Cage

Temporary Repair,

Valve

1BD 27

AFW Check Valve Backleakage

Snubber

Reduction

Program

1B

NSW Pump

TDH

Flow Trip Essential

Chiller

CSAS Test Logic

Steam

Hammer

SI and Containment Isolation Switches

Blowdown Time Delay Relay

Flow Switch Setpoint

AFW Isolation Conflict

1C Blowdown Valve Evaluation

1A blowdown Valve Evaluation.

Engineering

evaluations,

written

as

and also contained

in

as

engineering justifications,

appear

to be adequate.

The engineer-

ing evaluation

program for

10 CFR 50.59,

changes,

tests

and experi-

ments,

was

recently

modified

due

to

a

Inspection

Violation.

The present

evaluations

document

why the

issue "is" or

"is not"

an

unreviewed

safety

question

and

the limited number of

recent evaluations

reviewed

and available

appeared

to be adequate.

Temporary modifications

were

reviewed

and

the inspector

noted that

PCR 947, which expired

on June

6,

1987,

was not renewed until June

9,

1987.

The late renewal

was considered

to be

an isolated

admini stra-

tive error.

Modification PCRs

1286,

2292,

and

1391 were walked

down in the field.

Minor deviations

were noted in the field placement

of the

AFW check

valves

in

PCR 1286.

The seismic

analyses

for the five installations

of

1286 were reviewed

and were found to be acceptable.

1286,

2292

and

1391

were

reviewed

in detail.

The inspector

discussed

the various discrepancies

with the appropriate

engineering

personnel.

Two violations of requirements

were

noted

during

the

course of the inspection.

The

licensee

is

committed

to

Regulatory

Guide 1.64,

Revision 2,

(}uality Assurance

Requirements

for

the

Design

of

Nuclear

Power

Plants,

which

endorses

ANSI

N45.2.11-1974Property "ANSI code" (as page type) with input value "ANSIN45.2.11-1974" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,

guality

Assurance

Requirements

for the Design of Nuclear

Power Plants.

Section 3.1 of

this standard

states

the following:

Applicable

design

inputs,

such

as

design

bases,

regulatory

requirements,

codes

and

standards,

shall

be identified,

docu-

mented

and their selection

reviewed

and approved.

Changes

from

specified design

inputs,

including the

reasons

for the

changes

shall

be identified, approved,

documented

and controlled.

The design input shall

be specified

on

a timely basis

and to the

level of detail

necessary

to permit the design activity to

be

carried

out

in

a correct

manner

and to provide

a consistent

basis

for making design decisions,

accomplishing

design verifi-

cation measures,

and evaluating

design

changes.

The

intent

of this

standard

was

not

met for corporate

or site

generated

modification packages.

The modification procedure

MOD 204,

Modification

Implementation,

Revision 0,

does

not

specifically

require that design input be identified,

documented

and their selec-

tion

reviewed

and

approved.

None

of the

reviewed

contained

information formally identified as

design

input.

The various

contained

a collection of statements

and information,

but did not

offer

a

formal distinction

between

design

inputs,

assumptions,

speculations,

useful

information,

etc.

The

intent

of

the

ANSI

standard

is that the design

input

be officially identified

as

such.

The

requirement

that

the

design

input selection

be

reviewed

and

approved

was

not met.

The

only review afforded

the

design

input

selection

was

the

design verification, which is

an overall

design

package

review.

The design

input review and approval

required

by the

ANSI standard

were not performed.

The lack of adequate

design

input led to the following modification

problems:

PCR 2292

was to modify one

channel

of the auxiliary feedwater

(AFW)

isolation

logic.

It did

not

identify all

of

the

schematics

for circuit boards

0834,

0843,

and 0841.

This led to

a'ield

revision

of

the

original

modification

because

the

proposed circuit board contact

points

were already

being

used.

PCR 2292 also did not identify all of the schematics

for circuit

boards

0835,

0842,

and

0844.

This led to

a field revision of

the original modification in order to permit proper test switch

functioning.

1286 modified the

AFW discharge

piping.

Additional check

valves

were installed

and various

system hydrostatic tests

were

required.

The check valves

were installed in different sections

of the

AFW system which had different design

pressures.

The

did not identify the different design

pressures

for the

AFW

system.

While observing

the

pressure

test

in the field, the

Senior Resident

Inspector

informed the

licensee

that the

wrong

design

pressures

were

recorded

in the post modifica ion test

procedure.

The hydrostatic test

pressures

which were

based

on

system

design

were

also

recorded

in error.

The

recorded

pressures

would

have

over

pressurized

sections

of

the

AFW

system.

The correct

hydrostatic

pressures

were

used

in the

post modification test.

In all cases,

a complete reference

was not provided concerning

design

bases,

regulatory

requirements,

codes,

and

standards

used in formu-

latingg

the finished design product.

In general,

the

licensee

failed

to positively control

the

use

of design

inputs for their design

change

program.

This is contrary to ANSI N45.2. 11-1974,

Section

3. 1.

As

stated

previously,

the

licensee

endorses

ANSI

N45.2. 11-1974Property "ANSI code" (as page type) with input value "ANSIN45.2. 11-1974" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..

Section

4. 1, Design Process,

states

the following:

Design

activities

shall

be

prescribed

and

accomplished

in

accordance

with procedures

of

a type sufficient to assure

that

applicable

design

inputs are correctly translated

into specifi-

cations,

drawings,

procedures

or instructions.

The intent of this standard

was not met for modification packages,

PCR 2292

and

PCR 1391.

The design

packages

did not contain suffi-

cient

information

for

correctly

translating

the

design

into

instructions.

1391,

Snubber

Reduction,

was written in

1986

and detailed

the

process

for removing

snubbers

on the

excess

letdown line from the

reactor

coolant

system.

The

seismic

analysis

required that certain

struts

be

set

at specific

angles

to insure

continued

operability

after

system

heatup

and

seismic

event.

The modification required

that the specific strut angles

be within + 1.5 degrees

and others to

be within

+

1 degree.

Discussions

with engineering

disclosed

that

strut

angles

with

a tight tolerance

would have to

be

measured

and

verified using

a surveyor's

instrument,

a theodolite.

The

angles

were actually

measured

using

two squares.

Oue to the

design

and locations of the struts,

a highly accurate

rea'ding

could

not be obtained.

Depending

upon which side of the strut the measure-

ments were taken,

the

angle

could vary

as

much

as

several

degrees.

The method

used to determine

the strut angles

was not accurate

enough

to ensure

the accuracy

required

by the original modification package.

The modification package

did not adequately detail

the design

process

with the appropriate

instructions

for obtaining

the correct

angle.

The

design

activity

was

not

properly

translated

into

adequate

instructions.

0

J

PCR 2292,

AFW Isolation

Logic,

was

written to

change fail

open

contacts

to fail closed

contacts.

The

implementation

instructions

called for removal of primary

and

backup

fuses

in the related

card

rack.

The instructions

stated

"Ensure that

no other associated

loops

in protection

sets I, II, III, are

in test

or trip mode."

The

instructions

did

not

specifically identify all of the trips or

controls that were affected

when the fuses

were

removed.

A total of

23 channels

of instrumentation

were affected,

including P447, first

stage

turbine pressure,

which was

a

1 out of

2 logic circuit

and

caused

a reactor trip when it was deenergized.

The design instruc-

tions relied

on maintenance

and operation

personnel ability to verify

the

fused circuits

and to prevent

an actuation.

This is inappro-

priate for this

type of modification

which

deals

with multiple

electronic circuits.

The modification

package

did

not adequately

detail

the

design

process

with the

appropriate

instructions

for

preventing

a reactor trip or other actuation.

In

general,

the

licensee

failed

to

ensure

an

adequate

design,

including

design

instructions,

which

led

to

the

two identified

problems.

This is contrary to ANSI N45.2. 11-1974,

Section

4. 1.

Together,

these

examples

constitute violation 400/87-38-02,

Failure

to follow ANSI N45.2. 11-1974.

Sections

3. 1 and 4. 1.

During the inspection,

the inspector

noted that the licensee

appeared

to have extensive

design basis

documentation.

Specifically,

documen-

tation for seismic

supports

detailed

the original

design

and

the

individual revisions

of that

design.

The original design

and its

revisions

were

considered

QA records

and

were

maintained

as

QA

records.

The

small

sample

of records

reviewed

indicated that the

design

basis

documentation

for the Harris facility was excellen

The design

change

process

was discussed

in detail with the engineer-

ing staff.

With a few exceptions,

the design

process

appeared

to be

adequate

to

control

and

document

plant

changes.

More

complex

modifications

required

additional

design

documentation'nd

the

licensee

has taken

steps

to include more detail in the design

package

by revising

the

design

implementation

procedure.

The

engineering

staff was responsive

to the inspector's

various concerns.

The

assessment

of the

licensee's

program

to

implement

changes

to

existing plant configurations

was that the licensee

appeared

to

be

adequately

controlling plant changes,

with exception

to the

noted

deficiencies.

Quality Assurance/Quality

Control

Review of the Harris Nuclear Plant

(HNP)

QA organization effective-

ness

consisted

of an evaluation of both the site's

QA/QC organization

and the corporate's

Performance

Evaluation Unit (PEU).

Assessment

of

each unit's activities

was

conducted

by interviews with each

unit

supervisor.

A detailed

review

of audit/surveillance

schedules,

0

10

scheduling

compliance,

audit/surveillance

findings,

other

plant

identified

nonconformances,

adequacy

and

timeliness

of corrective

actions

and the project's

trend analysis

programs

was conducted.

The

evaluation

concluded

that

the

QA program is adequately

accom-

plishing its assigned

function of identifying

and correcting

site

problems.

This conclusion

is

based

on the following observation,

discussions,

and documentation

reviews:

The

audit/surveillance

unit

size

and

experience

level

is

adequate.

The

inspector

examined

the qualifications

of ten site

QA/QC

personnel

and six

PEU personnel

and determined

that they were

capable

in their

areas

of audit/surveillance

exper ti se.

Site

surveillance

has,

on occasion,

used

QA engineers

and

the

PEU

maintains

a current listing of technical

specialist

candidates

it uses

for specific

technically

oriented

audit

areas.

surveillance

personnel

have

been

broadening

and improving their

technical

surveillance

backgrounds

by receiving

cross-training

in IKC, maintenance

and

EERC fields.

Likewise, discussions

with

QA management

revealed

they were in the process

of standardizing

a method to improve the audit skills of the technical

auditors.

A

formalized

audit/surveillance

system

was

in

place

and

adherence

to schedules

was adequate.

Operation

Quality Assurance

Instruction

OQAI -50,

Surveillance

Scheduling,

and

Corporate

Quality

Assurance

Departmental

Procedure

CQAD -80-1,

Procedure

for

Corporate

QA

Audits,

requires

formalized

systems.

The

inspector

examined

the

audit/surveillance

planning

and

scheduling

matrices

for

1986-1987

and found them satisfactory

Audits/surveillances

conducted

appear

to

be satisfactory

in

depth

and

scope,

and they identify

some relatively significant

problems for management

corrective action.

The inspector

reviewed

the following audits/survei llances

and their

respective

checklists that were performed at

HNP during 1986-1987:

Audit/Surveillance

No.

Activit

Examined

QAA/0022-86-01

QAA/0022-87-05

QAA/0100-86-05

QAA/0100-87-01

QAA/0128-86-02

QAA/0128-87-01

QAA/0128-87-02

First Operations

QA Audit at

Operations -

HNP Engineering

(HPES) (Pining Support/

Restraint

Program)

HPES (Design Control/PCR

Program)

HNP Operations

QA/QC Unit

HNP Operations

QA/QC Unit

HNP Operations

QA/QC Unit

0

QAA/0162-87-01

86-316

87-030

87-038

87-062

87-065

87-070

87-137

87-145

87-173

87-174

Audit of Material Quality Section

Fuel Load/Technical

Support

PMU - Grout Inspection

and Testing

PMU Construction

Welding Activities

Electrical Separation

ISI - Pre-service

Inspection

Control

Room Activities Operation

Storage of Materials

PCR Design

and Verification by

SEU

Rework of Valve BD-VllSA

E6RC Shift Turnover

The inspector

observed

that the audit reports

and related checklist

content

appear

to

improve with the

latest

audits

conducted.

The

specifics

as

to

what

was

examined,

how

examined,

sample

size

considered,

accept/reject criteria used

and the acceptability of the

audited

item are

now routine checklist

comments.

It appears

that the

majority of the most significant problems

are being identified by the

site surveillance

group.

The site surveillance

group performs about

25 percent

performance

based

type

inspections

(identifying safety-

problems)

whereas

the majority of audits

to date

have

been primarily of the

compliance

nature.

Discussions

with the

PEU supervisor

revealed

CPKL's

new audit philosophy is changing

to

the performance

based

concept.

Discrepancies

identified by either site

QA/QC, corporate

audits

or

plant personnel

receive timely, appropriate

corrective action.

The inspector

reviewed

the following Nonconformance

Reports

(NCRs),

Nonconforming

Field

Reports

(NFRs - for non-significant

noncon-

formances),

and

Audit Deficiency

Reports

(ADRs) for the

above

attributes:

NCR/NFR/ADR No.

Titie

NCR OP-86-0057

NCR OP-86-0066

NCR OP-86-0068

NCR OP-86-0073

NCR OP-86-0089

NCR OP-86-0173

NCR OP-86-0183

NCR OP-86-0185

NCR OP-87-007

NCR OP-87-008

NCR OP-87-020

NCR OP-87-030

NCR OP-87-031

Hylomar Thread Sealer

Lack of

Comprehensive

Program

for

Vendor

Work

Modes

3 and

4

LOCA

Violation of OQAI-30

ASME Section III Code Class

2,

NC4500

Damaged

Safety

Flex

Meld Stress

Calculations

Design Control

For

NEM Support

Beams

Procedure

Violation

Whip Restraint Calculations

Improper Quality Class of PCR-592

Diesel

FSAR Commitment

Temporary Cable Separation

Violation

0

12

"NCR 85-0527

"NCR 85-0836

NCR 86-0011

NFR 87-009

NFR 86-002

NFR 87-011

NFR 87-050

ADR 0128-86-02-C 1

ADR 0128-86-02-C2

ADR 0128-86-02-C4

ADR 0162-87-01-Cj

ADR 0)62-87-Ol-C2

ADR 0100-86-05-Cj

ADR 0022-86-01-F2

ADR 0022-86-01-Cj

ADR 0022-87-05-C4

ADR 0022-87-05-C5

ADR 0022-87-05-C7

Containment

Spray

Pumps

Unit 2

Discrepancy

Between

Vendor

Manuals

and

As-built Condition - Unit 2

Diesel

Generator

Engine

Control

Panels-

Unit 2

PCR-842

Implementation Without Authorization

Records

Vault Failure to Meet Temperature-

Humidity Requirements

Control

Room Personnel

ID Violation

Control of Calibrated Tools

Failure to Identify and Correct Electrical

Separation

Problem

Failure to Control

Gauges

Failure to Document Inspection Activities

Identification

on

Two

Radiographs

Not

Legible

g Off-the-Shelf Storage

Requirements

PCR Design

Package

Missing

PSAR Figures

Three Student

Test

Records

Not Acceptable

Annual Retraining of

PNSC

Members

Annual

ALARA Report Overdue

Radiation Control Project Information Review

bv ALARA Subcommittee

Overdue

Certain

Plant

Procedures

Did Not Receive

an

ALARA Review

"Denotes

old construction

NCRs remaining

open to date which generally

involve Unit 2 equipment

being considered

for replacement

parts for

Unit l.

Examination of those

discrepancies

(NCRs,

NFRs,

or ADRs) tha

were

already

closed

out

in

the

system

revealed

they

appeared

to

be

properly

handled.

Satisfactory

corrective

actions

were

specified

and

the closeout

of the

subject

discrepancies

were

accomplished

by

reinspection/verification

of details as, necessary.

However,

numerous

closed

NCRs

had

been

granted extensions.

Because

of this, particular

attention

was devoted

to the

open

discrepancy

backlog

(NCRs,

NFRs,

ADRs).

Several

of these

were

reviewed

and discussed

in detail with

responsible

plant

personnel

to determine if these

items

were

in

process

for resolution

and that

long-term

items

were

not

due

to

inattention

by management.

As of October 31,

1987,

HNP had

a total of 53 open

NCRs.

This number

by itself is not alarming;

however,

examination of the monthly to al

of open

NCR trend since June

1987, indicates

a continuing increasing

open

number.

Additionally, 39 of these

53

NCRs have

had extensions

(many with multiple extensions)

that if continued

could result

in

a

unmanageable

backlog.

Apparently

the

site

has

recognized

this

potential

problem in that the

HNP Vice President

issued

a memorandum

to all plant managers

(dated August 21,

1987) requiring all extension

13

requests

to receive

his

concurrence.

Actions of this

type

give

credence

to active

management

participation

and

plant

involvement

preventing

a potential

problem from getting 'out of hand.

No similar

adverse

trends

were noted for NFRs.

The

PEU

has

performed

well in following up

and closing out ADRs.

Audits include previous

open

items

in their audit

scope

and either

close

the

item or provide

a status

update.

Four relatively minor

audit findings were

open at the close

of this inspection.

Respon-

siveness

to

ADRs

has

been

excellent.

Only

one

instance

was

identified

where

the

inspector felt

an

audit

response

was

weak,

in

that, it did

not

appropriately -address

the

cause,

effect,

corrective action,

or action to prevent recurrence

(0022-87-05-C4).

Mechanisms

wer e in place

to recognize

and prevent recurring or

repetitive discrepant

conditions

and

upper

management

was

made

aware of these

trends.

The Senior Executive Vice President

reviews

and signs

each corporate

audit report issued.

The

QA

Department

uses

several

excellent

nonconforming

trending/

status

programs

that

help identify adverse

trends

and

recurring

discrepant

conditions that fall with in thei r area of responsibility.

The inspector

examined

the following current

QA Department

di screp-

ancy

trend/status

reports

and

meeting

agenda

content

that

are

routinely presented

to upper level

management:

Corporate Quarterly Nonconformance

Trend Reports

(ADRs,

NCRs)

HNP QA/QC Monthly Reports

(NCRs,

NFRs)

HNP Review Meetings