ML17345A314

backup

supply.

system

were

installed under

PC/M 85-175,

Change

Request

No.

2, Nitrogen Station

Additions and Relocation

Unit 3, dated

March

19,

1987,

and

PC/M

85-176,

Change

Request

No.

4,

Nitrogen

Station

Additions

and

Relocation - Unit 4, dated

March 19,

1987.

The licensee

has revised

ONOP 0208. 11, dated

March 28,

1988, to reflect current setpoints

and

required

operator

action.

Upon receipt

of

a valid low pressure

alarm, the operators

are directed to valve in additional bottles of

and valve out the nitrogen bottles previously in operation.

The directions also specify the time frame within which the operator

must

take

action

to

ensure

a

continued

supply of nitrogen for

operation of the auxiliary feedwater

(AFW) control valves.

The inspector

reviewed administrative

procedure

(AP) 0190. 15, Plant

Changes

and

Modifications,

dated

June

14,

1988.

This

procedure

provides

specific

guidance

to

the

appropriate

plant

department

coordinator

and

requires

that training briefs,

and

operating

and

maintenance

procedure

revisions,

be

prepared

and

completed prior to

the turnover of

a

PC/M to the plant operating staff.

This item is

closed,

and also closes part of violation 250,251/86-26-05.

(Closed)

EA 86-20, I.A.2: Lack of Operator Training and Inadequacies

of Emergency

Operating

Procedure

(EOP) for the

Rupture of

a

Steam

Generator

Tube.

Inspection

Report

250,251/85-40

identified

a

safety

concern

pertaining to the ability of licensed operators

to isolate

steam flow

paths

from

a faulted

steam

generator

in the event of tube rupture.

During the inspection, it was determined that

AFW steam

supply valves

could not

be remotely

shut with an

AFW initiation signal

present.

These valves

were designed

to cycle

open during the event,

even if

the control

room hand switches

were held in the closed position.

The

licensee

failed to recognize this design

feature

and therefore

had

not

supplied

adequate

procedures

or sufficient training

on

an

alternate

means

of isolation

.

This issue

was significant in that

without an alternate

means

of isolation,

an unnecessary

radioactive

release

could have occurred following a steam generator

tube rupture

In response

to this item,

the current revision of

EOP 3/4-EOP-E.3,

Steam

Generator

Tube

Rupture,

dated

March 25,

1988,

addresses

the

inability to close,

from the control

room,

the

steam

supply motor

operated

valve

(MOV) from the

ruptured

steam

generator.

If the

reactor

control operator

(RCO) is unable

to close

the

steam

supply

MOV from the cont~ol

room,

the

EOP directs

the

RCO to dispatch

an

equipment operator to locally open the

steam

supply

MOV breaker

and

then manually close the valve. If the

steam

supply

MOV can

be closed

from the

control

room,

the

EOP directs

the

RCO

to

dispatch

an

equipment operator to locally open the

steam

supply

MOV breaker.

The

licensee

has

determined

that if the

steam

supply

MOV can

be

closed

from the control

room, that the equipment operator

should also

verify the valve is in the closed position after opening

the

steam

supply

NQV breaker.

The licensee

has

issued

night orders

informing

the operating staff of this additional

information and

has committed

to formally incorporate

the verification of the valve position after

opening the

steam

supply breaker in EOP-E.3

by October

1,

1988.

This

item is closed,

and also closes part of violation 250,251/86-26-05.

(Closed)

EA 86-20, I.A.3: Inadequate

Operator Training and Incorrect

Procedural

Information.

Inspection

Report 250,251/85-40 identified

a failure by the licensee

to provide licensed operators

with training that 'specifies

the time

available

to

the

RCOs

to

take

action

following

a

low pressure

alarm.

The

value

provided

to the

operators

in training

differed

from that specified

in operations

procedures

and

in the

system description.

As

noted

in item I.A.1, the licensee

has

issued

PC/Ms85-175

and

85-176, to provide enhanced

system operability.

Review of the system

description,

training materials,

drawings,

and

procedures

indicates

setpoint

agreement

with the

PC/M.

AP 0190. 15 assigns

responsibility

for plant design

and control,

and maintenance

of design

documents,

to

the

power plant

engineer.

In addition

to the

requirements

of

AP

0190. 15,

the

1 i cen see

uti l izes

JPE-gI

3. 1,

Contr ol

of

Desi gn

Performed

by JPE,

dated June

23,

1987, to ensure that

PC/Ms generated

by

the

licensee's

Power

Plant

Engineering

Department

adequately

control

the

design

efforts for structures,

systems

and

components

important to nuclear

safety.

This item is closed,

and also closes

part of violation 250,2Sl/86-26-0S.

(Closed)

EA 86-20, I.A.4: Failure to Provide Adequate

Procedures

Inspection

Report

250,251/85-40

delineated

the

failure

by

the

licensee

to

provide

adequate

emergency

operating

procedures.

Specifically,

emergency

operating

procedures

did not give sufficient

guidance

to assure

that the required

AFW flow is supplied to each

unit within three minutes,

in the event of a two unit trip with only

one

AFW pump available.

To resolve this concern,

the licensee

has included specific direction

in 3/4-EOP-ECA 0.0,

Loss of All A.C. Power,

dated April 16,

1987.

A

caution statement

has

been included to direct operations that, in the

event both units require

AFW under natural circulation conditions

and

trai n

1

of

AFW i s

inoperabl e,

within

3

minutes

the

AFW f1 ow

controllers

should

be placed

in manual

and adjusted

to

300

gpm per

unit.

Licensed Operator Instructor

Lesson

Plan

No. 0063-0L, Appendix

00,

ECA 0.0,

dated July 9,

1986, requires

the instructor to discuss

the basis for all cautions

and notes in the procedure.

Additionally,

the

lesson

plan

handouts

provide

an

explanation

of the setpoints

provided in the procedure.

This item is closed,

and also closes part

of violation 250,251/86-26-05.

(Closed)

EA 86-20,

I.A.5: Failure to Provide

Adequate

Procedures

Covering

a Safety Related Activity.

Inspection

Report 250,251/85-40 delineated

a fai lure by the licensee

to provide

adequate

procedures

to cover

safety related activities.

Specifically, procedures

did not specify. local operation of'rain

2

of AFW when 'the main control

room is inaccessible

and only the

B AFW

pump is available.

Additionally, instructions

were not provided

on

how to locally reset

and restart

a tripped

AFW pump.

To

resolve

these

concerns

the

licensee

has

revised

O-ONOP-103,

Control

Room Inaccessibility,

dated July 18,

1987,

to direct local

operation

of the

AFW system,

provide provisions for operation

of

train

1 or

2 of

AFW and

provide

instruction

on

the

process

for

resetting

and

restarting

a

tripped

AFW

pump.

Additionally,

Inspection

Report

250,251/87-07

documented

observations

of

a walk

through

of

the

aforementioned

procedure.

Several

areas

of

improvement

were

noted including reset of

a tripped

AFW

pump

and

local

operation

of the

AFW system.

This item is closed,

and also

closes part of violation 250,251/86-26-05.

(Closed)

EA 86-20, I.B: Failure to Correctly Translate

Design Inputs

into Plant Drawings.

Inspection

Report 250,251/85-40

documents

the licensee's

failure to

correctly translate

design inputs into plant drawings.

Specifically,

the nitrogen

backup

system

drawings incorrectly identified pressure

control valves

as being set at

55 psig versus

the correct value of 80

psig.

The setpoint

change

was brought about during implementation of

the

PC/M.

To resolve this

concern

the licensee

has

revised

Instrument

Index

Sheet,

5610-M311,

Pressure

Controllers,

dated

September

18,

1986 and

Piping

and Instrument

drawing

5610-M339, Auxiliary Feedwater,

Main

Steam Isolation Valve,

and Pressurizer

PORV Nitrogen

Backup

Supply

Systems,

dated April 26,

1988.

These revisions correctly reflect the

value of 80 psig for the nitrogen pressure

control valves.

As noted

in item I.A.3 above,

the

power

plant

engineer

is

responsible

for

plant design

and control,

and the maintenance

of design

documents.

This

item is

closed,

and

also

closes

part

of violation

250,

251/86-26-05.

(Closed)

EA

86-20,

I.C.1-2:

Failure

to

Correctly

Translate

Appropriate Quality Standards

into Procedures

or Drawings.

Inspection

Report 250,251/85-40

delineated

a failure by the licensee

to properly co'ntrol

design

changes

in that required

changes

to the

licensee's Q-list were not completed following completion of a

PC/M.

This error resulted

in components

of the nitrogen backup

system not

being maintained

commensurate

with their function.

To

resolve

this

concern,

the

licensee

has

in

place

Quality

Instruction

JPE-QI

2.7,

Nuclear

Plant Q-List,

dated

November

30,

1987,

which sets

forth the requirements

that the Q-list be revised

for impact from:

(1)

PC/Ms;

(2)

New or revised

drawings

resulting

from plant design

changes,

plant modifications, resolution of nonconformances,

and drawing

discrepancy

resolution;

(3)

Technical Specification

(TS) changes;

and

(4)

Correspondence

between

the licensee

and the

NRC.

Review of the Q-List for'tems85-176

revealed

the

correct

classification.

Purchase

orders

materials

were

purchased

to the

item

is

closed,

and

also

250,251/86-26-05.

associated

with

PC/Ms85-175

and

incorporation

of

the

safety

were also

reviewed to ensure

that

appropriate

classifications.

This

closes

part

of

violation

(Closed)

EA 86-20, I.C.2: Safety Related Application.

Inspection

Report

250,251/85-40

delineated

the

failure

by

the

licensee

to

provide

the

required

protection

against

commo'n

mode

failure for the nitrogen

backup

low pressure

alarm.

To resolve this issue,

the

licensee

has

completed

PC/Ms85-176

and

85-175,

which installed

safety

grade

pressure

switches

along with

supporting

equipment

and

instrumentation.

Review

of

drawing

5610-E-27,

SH.34,

Elementary

Diagram

N2 Backup Supply Stations

1 & 4,

Low Pressure

Alarms, dated January

1,

1988,

and associated

referenced

drawings,

confirmed

the

changes

and the protection

against

common

mode failure for the nitrogen

backup

low pressure

alarm.

This item

is closed,

and also closes part of violation 250,251/86-26-05.

(Closed)

EA 86-20,

I.D.1-4:

Failure to Correctly Translate

Design

Inputs into System Descriptions or Design Basis

Documents.

Inspection

Report

250,251/85-40

documents

the licensee's

failure to

correctly translate

design

inputs into the

system

description

or

design basis

documents.

I

Review of System Description

No. 0709117, Auxiliary Feedwater

System,

dated

April 22,

1988,

reflects

the

correct

incorporation

of the

backup

operating

pressure

of

80 psig,

container

backup

low pressure

alarm setpoint of 650 psig, accurate

description

of the train configuration

and

the

time associated

with operator

action

in order to preclude

loss of the nitrogen

backup.

As noted

above,

the licensee

has

adequate

procedures

in place

to ensure

that

the appropriate

plant documentation

is updated

when

PC/Ms are issued.

This

item is

closed,

and

also

closes

part

of violation

250,

251/86-26-05.

(Closed)

EA 86-20,

I.E:

Failure to Evaluate

the

Impact of Design

Changes

on the

AFW System.

Inspection

Report

250,251/85-40

delineated

the licensee's

failure to

evaluate

the

impact of design

changes

on the nitrogen

consumption

rate

by the

AFW flow control valves

and the time period allowed for

operators

to valve in additional nitrogen bottles.

To

resolve

this

item,

the

licensee

has

developed

operational

surveillance

procedure

3/4-0SP-075.3,

AFW Nitrogen Backup

System

Low

Pressure

Alarm Setpoint,

dated

May 5,

1988, 3/4-0SP-075.6,

Auxiliary

Feedwater

Train

1

Backup Nitrogen Test,

dated

March

28,

1988

and

3/4-0SP-075.7,

Train

2

Backup

Nitrogen Test,

dated

March

28,

1988.

These

procedures

provide for the

dynamic

testing

of the nitrogen

consumption

rate

and the low pressure

alarm

setpoint for both trains

1

and

2 of the

AFW backup

system

flow control

valves.

This

item

is

closed,

and

also

closes

part of violation 250,251/86-26-05.

(Closed)

EA 86-20, I.F:

Failure to Adequately

Document Assumptions

and Design Inputs in Calculations for Plant Modifications.

Inspection

Report

250,251/85-40

documents 'the licensee's

failure to

adequately

document

assumptions

and

design

inputs

in calculations

used for plant modifications.

To

resolve

this

item,

the

licensee

has

in

place

Administative

Procedure

JPE-AP3.9,

Calculation Standard

Format, dated

September

10,

1986,

which

requires

that calculation

packages

be

completed

and

assumptions

be

clearly

stated

such

that

no

additional

verbal

explanation

is required,

and that all basic data

and assumptions

be

provided with references

or basic

discussions.

Additionally, the

pagination/revision

scheme

for the calculation

package facilitates

a

cursory review for original order, missing or out of sequence

pages,

and the correct revision of each

page.

This item is closed,

and also

closes part of violation 250,251/86-26-05.

Escalated

Enforcement

Action (EA) 86-20-IV:

Failure to Establish or

Implement Adequate

Procedures,

and to Properly Control the Revision

and Distribution of Safety-Related

Procedures.

This

EA originated

from violation

250,

251/86-26-09,

which

encompassed

previously

identified unresolved

items 250,251/85-40-03,

250,251/85-40-17,

and

250,251/85-40-20.

Status

Item

EA 86-20,

IV.A.1

EA 86-20,

IV.A.2

EA 86-20,

IV.B

closed

cl osed

The basis for closure of each

item is discussed

below.

(Closed)

EA

86-20,

IV.A.1:

Failure

to

Establish

Procedures

Specifying Independent

Verification for Safety Related

Systems.

Inspection

Report

250,251/85-40

delineated

the licensee'

fai lure to

provide

and implement adequate

procedures

to ensure that independent

verification was performed

and documented

on the return to service of

instrumentation vital to the operation of the auxiliary feedwater

and

backup nitrogen

systems.

The licensee

has in place

O-ADM-031, Independent Verification, dated

April

26,

1988,

which

specifies

the

overall

plant

policy for

independent verification and delineates

those safety related

systems

which require

independent

verification,

including

the

AFW system.

This

item is

closed,

and

also

closes

part

of violation

250,

251/86-26-09.

(Cl osed)

EA 86-20,

IV.A.2:

Fai 1 ure

by the

P 1 ant

Nuc1 ear

Sa fety

Committee

to

Review

Nuclear

Safety-Related

Temporary

System

Alterati on s.

Inspection

Report

250,251/85-40

delineated

the

licensee's

Plant

Nuclear Safety Committee'

(PNSC) failure to review Temporary

System

Alterations

(TSAs)

3-84-11-75,

3/4-85-8-75,

3/4-84-99-75,

and

3/4-84-100-75 within fourteen

days of the

Plant

Supervisor-Nuclear

approval date.

The inspector

reviewed current

TSAs to confirm review by the

PNSC

within the required

time frame.

No discrepancies

were identified.

The inspector's

review of O-ADM-503, Control

and

Use of Temporary

System Alterations, dated June 2,

1988,

noted that

TSAs for inservice

equipment

may be installed without prior PNSC approval,

based

on the

determination

by the Shift Technical Advisor and the Plant Supervisor

Nuclear that

the

TSA is identical

or similar to

one previously

installed.

(TSAs installed

under this

method

are still required to

be

reviewed

by

the

PNSC within the

required

time

frame).

The

procedure,

however,

does not provide guidance

on the determination of

identical

or similar.

To rectify this situation,

the licensee

has

committed to remove the possibility of TSA installation

based

on

an

identical

or

similar determination

from the

procedure

currently

10

undergoing

revision.

This item is closed,

and also

closes

part of

viol ati on 250,251/86-26-09.

(Closed)

EA 86-20,

IY.B:

Failure to Ensure that

a Safety-Related

Procedure

was Properly Controlled

Inspection

Report 250,251/85-40

delineated

the licensee'

failure to

ensure

that

a safety-related

procedure

was

approved

for release

by

authorized

personnel

and

appropriately

distributed

prior to

the

cancellation of the previous procedure.

Review of ADM-0109. 1,

Preparation,

Revision,

Approval,

and

Use of

Procedures,

dated

dune

7,

1988,

indicates

that the licensee

has

in

place

adequate

controls

to

ensure

that

and

commitments

are

addressed

prior to the

issuance

of new or revised

procedures.

This item is closed,

and also closes part of violation

250,251/86-26-09.

3.

Followup

on

Unresolved

Items

(URIs),

Inspector

Followup

Items (IFIs),

Inspection

and

Enforcement

(IE)

Information

Notices, IE

Bulletins

(information only), IE Circulars

and

NRC Requests

(92701).

(Cl osed)

IFI

250,251/88-11-03

pertains

to

the

resolution

of

the

differences

in

documentation

associated

with the

ICW

gauge

assembly

material.

It has

been

determined

that the

requirements

of

10 CFR 50,

Appendix

B, Criterion III were violated

in that

a

change

to

a design

specification

was

made

without

appropriate

review

and

evaluation.

Corrective action will be tracked

along with violation 250,251/88-14-01,

which is di scussed

in paragraph

5 of this report.

This IFI is closed.

(Closed)

URI

250,251/85-40-29:

Inspector

Concerns

Pertaining

to

the

Sizing of Motor Thermal Overload Relays

(REF. 87-32,

paragraph 5.g).

Inspection

Report

250,251/85-40

delineated

a concern

pertaining to the

apparent

difference

between

the

manufacturer's

recommended

thermal

overload

size

and the size

chosen

by the licensee

to satisfy Regulatory

Guide

1. 106,

Thermal

Overload

Protection

for

Electric

Motors

on

Motor-Operated

Valves.

The licensee's

Power Plant Engineering

group has completed preparation

and

departmental

approval of Standard

No. E-3.20,

Motor Operated

Valve Thermal

Overload

Heater

Relay Selection

Criteria.

The licensee's

integrated

schedule

projects that the review of all safety related

motor operated

valve thermal

overload relays in accordance

with the selection criteria to

will be completed

by November

1988. This review will include the issuance

of the

appropriate

PC/Ms

as

necessary

to correct

design deficiencies.

This item is closed.

(Closed)

IFI

250,251/85-40-38:

Inspector

Concerns

Pertaining

to

Licensee's

Fail

Safe

Testing

of the Auxiliary Feedwater

Flow Control

Valves

(REF. 87-32,

paragraph 5.j).

Inspection

Report 250,251/85-40

delineated

a concern

over the licensee's

method of fail safe testing the

AFW flow control valves.

The

inspector

reviewed

OP

0209.1,

Valve

Exercising

Procedure,

dated

Nay 19,

1988;

3/4-OSP-075. 1,

Auxi 1 iary

Feedwater

Train

1

Oper abi 1 ity

Verification, dated

February

16,

1988;

drawing 5610-T-E-4061,

sheet

4,

B Emergency Diesel Generators

Pumps

Steam

Supply

Systems,

dated

November 28,

1987;

and drawing 5610-T-E-4062,

sheet

3,

Steam

Generator Auxiliary Feedwater

Supply Systems,

dated

November 28,

1987;

and confirmed that the

AFW flow

control

valves

are

tested

in

a fail

safe

manner

in accordance

with

licensee

commitments.

This item is closed.

(Cl osed)

URI

250,251/86-18-13:

Inspector

Concerns

Pertaining

to

the

Licensee's

Loss of DC Power Procedure

(REF. 87-32,

paragraph

5.o).

Inspection

Report 250,251/86-18

noted that the licensee's

Phase

I review

identified the absence

of a loss of

DC power procedure.

Additionally, the

Institute of Nuclear

Power Operations

issued Significant Operating

Event

Reports

(SOER)

81-15

and 83-5,

which document

industry events

involving

the failure of various aspects

of a plant's

DC power system.

Currently

the

licensee's

engineering

department

has -forwarded to the

Procedure

Upgrade

Program

(PUP)

group the engineering

evaluation

on the

loss

of

DC

bus

study.

The engineering

study analyzed

the

independent

failure of each vital

DC bus

and

addressed

the

recommendations

of the

SOERs.

The

study

addressed

SOER

recommendations

by identifying all

equipment

powered

from each vital

DC bus,

the fai lure

mode

on

a loss of

power,

and analysis

of the

systems

required for safe

shutdown for each

vital

DC bus failure.

The

PUP group will address

the study's findings via the generation

of four

off-normal

operating

procedures,

O-ONOP-003.14,

.15,

.16,

and

.17,

to

fully address

potential

loss of

DC

bus

scenarios.

Issuance

of these

procedures

is scheduled for October

1988. This item is closed.

(Open)

IFI 250,251/87-07-03:

Generation

of Electrical

Breaker Setpoint

Document

(REF. 87-32,

paragraph

5.u)

Inspection

Report

250,251/87-07

documents

the

l icensee'

request

for

engineering

assistance

for the

development

and

issuance

of

a breaker

setpoint document.

This item was

issued

in conjunction with the closure

of URI 86-37-01

and IFI 85-22-11,

which pertained to the establishment

of

DC feedbreaker

and inverter input breaker trip setpoint

settings'he

licensee's

integrated

schedule

currently projects modification 0618,

Add to Breaker List Breaker Trip Setpoints,

to start in June

1989 with

completion for December

1989.

An additional

aspect

of the

licensee's

electrical coordination documentation

includes modification 1254,

Add Fuse

Specifications

to Breaker List and Drawings,

scheduled

to start

in June

1989

and be completed in January

1990.

12

This item will remain

open

pending completion of a breaker trip setpoint

and fuse specification

document.

Monthly and Annual Surveillance

Observation

(61726/61700)

The

inspectors

. observed

TS required

surveillance

testing

and verified:

that

the test

procedure

conformed to the requirements

of the

TS; that

testing

was

performed

in accordance

with adequate

procedures;

that test

instrumentation

was calibrated;

that limiting conditions for operation

(LCO) were met; that test results

met acceptance

criteria requirements

and

were

reviewed

by personnel

other than the individual directing the test;

that deficiencies

were identified,

as

appropriate,

and

were

properly

reviewed

and resolved

by management

personnel;

and that system restoration

was adequate.

For completed tests,

the inspectors

verified that testing

frequencies

were met and tests

were performed by qualified individuals.

The

inspections

witnessed/reviewed

portions

of

the

following test

activities:

4-SMI-041. 16

TAVG/Delta T Protection

Channels

Periodic Test

4-0SP-072.2

Main

Steam

Isolation

Valve

(MSIV) Nitrogen

Backup

Periodic Test

No violations/deviations

were identified in the areas

inspected.

Maintenance

Observations

(62703/62700)

Station

mainten'ance

activities of safety related

systems

and

components

were

observed

and

reviewed

to ascertain

that

they

were

conducted

in

accordance

with approved

procedures,

regulatory guides,

industry codes

and

standards,

and in conformance with TS.

The following items

were considered

during this review,

as appropriate:

That

LCOs were met while components

or systems

were

removed

from service;

that approvals

were obtained prior to initiating work; that activities

were

accomplished

using

approved

procedures

and

were

inspected

as

applicable;

that procedures

used

were

adequate

to control the activity;

that

troubleshooting

activities

were

controlled

and

repair

records

accurately

reflected

the maintenance

performed; that functional testing

and/or

calibrations

were

performed

prior to returning

components

or

systems

to service; that

QC records

were maintained; that activities were

accomplished

by qualified personnel;

that parts

and materials

used

were

properly certified; that radiological controls were properly implemented;

that

QC hold points

were established

and

observed,

where required; that

fire prevention controls

were

implemented;. that outside

contractor

force

activities were controlled in accordance

with the approved

QA program;

and

that housekeeping

was actively pursued.

13

The

inspectors

witnessed/reviewed

portions of the following maintenance

activities in progress:

Repairs to

ICW Pump

3A and

3C Discharge

Pressure

Gauge Fittings.

Troubleshooting

Unit 3 Rod Position Indicator Alternate

Power Supply

Breaker

(LP 317-18).

Troubleshooting Unit 3,

3B Main Steam Isolation Valve Nitrogen Backup

Regulator Failure.

a.

ICW Pipe Fitting Design

Change

IFI 250,251/88-11-03,

identified conflicting requirements

in licensee

design specifications for ICW gauge

assembly fittings. The licensee's

original

and current design specifications,

5610-M-50 and 5177-PS-11,

require the

use of stainless

steel fittings for ICW gauge

assemblies.

However,

Non-conformance

Report

(NCR)86-112,

dated

March 5,

1986,

was

issued

to document

the

use of nonconforming

ICW pump discharge

pressure

gauges

and

piping.

Attachment

D to

the

NCR,

drawing

5610-J-155-P10,

note

12, allowed the

use of carbon

steel

for piping

between

the

root valve

and

the

main

header

for gauge

PI-3-1452.

Discussions with the licensee

indicated that the drawing in the

NCR

documented

the "as-found" condition of the

ICW gauges

and associated

piping,

and

erroneously

allowed

the

use

of carbon

steel

in this

portion of the system.

In addition,

the licensee

could not determine

when

the

nonconforming

material

was

installed

or

whether

any

engineering

evaluation

occurred prior to the material substitution.

10 CFR 50, 'Appendix B, Criterion III, as

implemented

by the approved

Florida

Power

and

Light

Company

Topical

Quality Assurance

Report

(FPLTQAR) 1-76A, Revision

11, Topical Quality Requirement

(TQR) 3.0,

Revision 5, requires that design

changes

be subject to design control

measures

commensurate

with those applied to the original design,

and

that these

design control measures

assure

that applicable

regulatory

requirements

and

the

design

basis

are correctly translated

into

specifications,

drawings,

procedures

and instructions.

FPLTQAR 1-76A,

Appendix

C,

Revision

7 specifically

commits,

with

exceptions

not

relevant

here,

to

American

National

Standards

Institute

(ANSI) N45.2. 11-1974,

Quality Assurance

Requirements

for

the

Design of Nuclear

Power Plants,

and to Regulatory

Guide 1.64,

Revision 2, Quality Assurance

Requirements

for the Design of Nuclear

Power Plants,

which endorses

ANSI N45.2. 11-1974.

ANSI N45.2.11-1974

specifies that

documented

procedures

shall

be

provided

for design

changes

to approved

design

documents

which assure

that the impact of

the change is carefully considered,

required actions

documented,

and

information

concerning

the

change

is transmitted

to all affected

persons

and

organizations.

These

changes

must

be justified

and

subjected

to control measures

commensurate

with those applied to the

original design.

14

Contrary to the above,

on March 5,

1986,

a change

was

made to the

ICW

design

specifications

through

the

resolution

of

non-conformance

report 86-112,

attachment

D, drawing

5610-J-155-P10,

note

12,

which

allowed the

use of carbon

steel

pipe fitting between

the root valve

and

the

main

header

for pressure

gauge

PI-3-1452.

The

increased

susceptibility of the

carbon

steel

to .corrosion

was

not carefully

considered

or

justified

through

an

engineering

evaluation.

Subsequently,

on April 27,

1988,

the fitting failed due to salt water

induced

corrosive

wear.

This resulted

in the

3A

ICW

pump

being

placed

out of service.

The failure to

meet

the

requirements

of

10 CFR 50, Appendix

B is

a violation (250,251/88-14-01).

b.

Rod Position Indication (RPI) Alternate

Power Supply Breaker Sizing

'I

~

I

On June

17,

1988,

at

1446, with Unit 3 at

100% power,

the breaker

supplying alternate

power to the

RPI circuitry tripped,

causing

a

loss of all

RPI for the unit.

The licensee

was performing Section

6. 5 of

3-PMI-028. 2,

Axial

Flux

Rod

Deviation

and

Rod

Position

Indication Monthly Test,

which allows power to the

RPI to be shifted

to the alternate

power

supply in lieu of the

RPI inverter.

After the

alternate

power

supply

was lost,

power

was

restored

to the

RPI

circuitry through the normal inverter.

This situation again occurred

at

0125

on

June

19,

1988,

when

the

procedure

was reinitiated

to

complete the testing.

The breaker

was reclosed

and the procedure

was

completed satisfactorily.

Troubleshooting

was initiated to determine

the

cause

of the

inadvertent

tripping of the

RPI alternate

power

supply breaker

(BKR LP 317-18).

Initial inspection

revealed that the

breaker

was

rated

at

20

amperes

and

load for the circuit was

approximately

21-22

amperes.

PC/M 82-121,

which installed

the

RPI

alternate

power supply,

indicated that the breaker

had

a

30

ampere

rating.

The licensee

obtained

a gL-1 replacement

breaker,

rated at

30 amperes,

and installed it in BKR LP 317 at position

18.

Initial

indication was that

PC/M 82-121

may have

been

inadequate

in that it

did not require replacing

the existing

20

ampere

breaker with the

required

30 ampere

breaker.

However, the inspectors

and the licensee

are still researching

the issue.

It will be addressed

in the next

report.

This issue is identified as

an inspector followup item (IFI

250,251/88-14-02).

6.

Operati ona1

Safety Verificati on (71707)

~

The inspectors

observed control

room operations,

reviewed applicable logs,

conducted

discussions

with control

room

operators,

observed

shift

turnovers

and

confirmed operability of instrumentation.

The inspectors

verified the operability of selected

emergency

systems,

verified that

maintenance

work orders

had

been

submitted

as required

and that followup

and prioritization of work was

accomplished.

The

inspectors

reviewed

tagout

records,

verified compliance with TS

LCOs and verified the return

to service of affected

components.

15

By observation

and direct interviews,

verification

was

made

that

the

physical security plan was being implemented.

Plant

housekeeping/cleanliness

conditions

and

implementation

of

radiological controls were observed.

Tours of the intake structure

and diesel, auxiliary, control

and turbine

buildings were

conducted

to observe

plant equipment conditions including

potential fire hazards,

fluid leaks

and excessive

vibrations.

The

inspectors

walked

down accessible

portions of the following safety

to verify operability

and proper valve/switch alignment:

A and

Control

Room Vertical Panels

and Safeguards

Racks

Intake Cooling Water Structure

4160 Volt Buses

and

480 Volt Load and Motor Control Centers

Unit 3 and

4 Feedwater

Platforms

Unit 3 and

4 Condensate

Storage

Tank Area

Area

Unit 3 and

4 Main Steam Platforms

No violations/deviations

were identified in the areas

inspected.

~

7.

Physical Security (71881)

Station security activities were observed

during this inspection period to

ascertain

that

they

were

conducted

in

compliance

with the

approved

Physical

Security Plan

(PSP).

The following attributes

were

considered

during these

observations,

as

appropriate:

that the minimum number of armed guards is

on site for each

shift; that

search

equipment

such

as x-ray machines,

metal detectors

and

explosives detectors

are operational;

that the protected

area

(PA) barrier

is well maintained

and is not compromised

by erosion,

opening in the fence

or walls, or proximity of vehicles or other objects that could be used to

scale

the barrier;

that illumination in the

PA is

adequate

to allow

patrolling guards

to observe

the

area

at night

and

permit the

use

of

closed circuit monitor s by alarm station

operators;

that the vital area

(VA) barriers

are well maintained;

that persons

granted

access

to the site

are

badged

to indicate whether

they

have

unescorted

or escorted

access

authorization; that there are

no obstructions

in the isolation

zone that

could

conceal

an

individual

attempting

an

unauthorized

entry

or

interference

with the detection/assessment

system;

and that

when

search

equipment

or alarm

systems

are inoperable,

or when there is

a breach of

the

PA or

VA barrier,

the licensee

implements

appropriate

compensatory

measures.

No violations/deviations

were identified in the areas

inspected.

16

8.

Reactor

Vessel

Pressure

Transient Protection (TI 2500/19)

The

purpose

of this

inspection

was

to verify that

the

licensee

has

implemented

commitments contained

in correspondence

Safety

Issue

A-26,

and the

safety evaluation

reports

concerning

reactor

vessel

pressure

transient protection.

The

items to

be verified have

been divided into several

areas:

design;

administrative

controls

and

procedures;

training

and

equipment

modification; and surveillance;

and are discussed

below.

a

~

Design

PC/M

75-81,

entitled Nil Ductility Transition

Temperature

(NDTT)

Contr ol,

was

impl emented

on January

6,

1978 for Unit

3

and

on

November

9,

1977 for Unit

4.

This

PC/M modified

PORV control

circuits to provide low pressure relief settings.

The setpoint of 415

psig for low temperature

operation

(below 300F)

is designed to

keep

the primary loop pressure

below the Appendix

G limits.

The following

PC/Ms were

subsequently

implemented

to

meet

the additional

design

requi rements:

PC/M 78-27,28,

Overpressure

Mitigation System

(OMS)

Permissive

Status

Panel

Light

and

Annunciator

Interlocks;

PC/M

78-16,17,

Pressurizer

PORV's

Backup

N-2 Supply;

and

PC/M 78-23,24,

OMS Test Switch and Relabel

OMS Components.

These

modifications

are

discussed

further in the following sections.

The pressurizer

PORVs

are

spring-loaded-closed.

Air is required to

open the valves

and is supplied

by instrument air.

In the event of a

loss of instrument air,

a

backup

(N-2) system is provided

which will supply

enough

N-2 for

a

minimum of ten

minutes

of

operation.

Drawing 5610-M-339

shows

the

N-2 backup

system provided

for each

PORV.

Each

PORV has

two redundant

solenoid valves which are

energized

in order to

open

the

PORVs.

These

solenoid

valves fail

closed

on

a loss of power.

However,

each solenoid is powered off the

125 volt vital

DC supply.

Therefore

on

a loss of offsite power, the

station batteries will be available to allow the operation

of the

PORV.

Drawing 5610-E-25,

sheet

64,

shows

the wiring configuration

and power supply for the

PORV solenoid valves.

There are two PORVs

and their associated

block valves which are

shown

on drawing 5610-T-E-4501.

If one

PORV is inoperable,

the remaining

PORV is capable of relieving the

RCS to prevent exceeding

Appendix

G

limits. If the

PORV fails in the

open position, its associated

block

valve

can

be closed to isolate

the

PORV to terminate

the pressure

Drawings

5610-E-855

and

5610-E-25,

sheet

27,

show the

wiring configuration

for

the

motor

operated

block

valve.

By

referencing

the breaker list, the inspector verified that the block

valves

were

powered

from

a separate

vital power source.

The block

valves

are

powered

from the vital portion of the

3B and

4B motor

control

centers

(MCC) for Unit

3

and

4 respectively.

Each

block

valve fails

as is.

The failure of one block valve will not affect

17

the operability of the associated

PORV, therefore it would not create

a pressure

Each

PORV is opened

by the energization

of two solenoid valves which

realign to allow instrument air or N-2 to flow to the

PORV actuator.

These

solenoids

are

redundant

such

that

the fai lure of

one will

prevent the

PORV from opening.

If the

PORV was

open

and the solenoid

valve failed,

the

PORV would fail closed.

However,

the remaining

PORV would not be affected

and could be used to mitigate the pressure

Drawing 5610-T-D-16A shows the control

system for the

OMS.

Each

PORV

has

its

own

switch

installed

on

the

main

control

board.

The

operator

can enable/disable

the

OMS by selecting

"LO Pressure

OPS" or

"Normal

OPS",

respectively.

The

setpoint

pressure

and

actual

pressure

are

derived

from

redundant

temperature

and

pressure

transmitters.

PORV

456,

which is the

primary

OMS channel,

uses

temperature

element

(TE)

430B

and

pressure

transmitter

(PT)

403.

PORV 455C, the backup

OMS channel,

uses

TE-423B and PT-405.

The

licensee's

Appendix

G curve

used

for overpressure

transient

analysis

shows the limit at 100F to be

510 psig.

In conjunction with

the Nuclear

Steam Supply

System

vendor,

the

PORV setpoint

overshoot

was determined

to

be

78 psig.

With

a relief setpoint of 415 psig,

the final pressure

of 493 psig is reached

for the worst

case

mass,

input transient.

This setpoint is acceptable

since it will keep the

pressure

below the Appendix

G limit of 510 psig.

The limiting heat

input cases

calculations

also

show

a maximum pressure

transient

below

that allowed by Appendix

G limits.

The inspector

reviewed

the licensee's

safety evaluations

which were

completed for each

PC/M associated

with the

OMS.

All the information

was complete.

Administrative Control

and Procedures.

No procedure exists to specifically minimize the time the reactor is

in a water-solid condition.

However,

when operating

in a water-solid

condition,

precautions

are

taken to minimize the possibility of

a

pressure

transient,

some of which are discussed

below.

To minimize the possibility of

a

pressure

transient

while in

a

water-solid condition, Operating

Procedure

OP-041.1, entitled Reactor

Coolant

Pump,

specifies

that

when the

RCS is solid,

a

RCP shall not

be started if the

steam

generator

blowdown temperature

exceeds

the

RCS temperature

by 10F.

This will prevent the Appendix

G limits for

the heat input case

from being exceeded.

General

Operation

Procedure

(GOP) 305, entitled

Hot Standby to Cold

Shutdown,

contains

instructions

to isolate

the hot leg and cold leg

safety

injection

valves

prior to

cooling

the

RCS

below

380F.

18

Attachment

1 to this procedure

also requires

the operators

to sign

off that the safety injection valves

are

closed

and their breakers

are locked open,

and to isolate the pressurizer

heaters

to reduce

the

possibility of a pressure

The operators

are alerted

to the automatic

operation

of the

OMS by

annunciators

provided

on vertical

panel

A in the control

room.

In

response

to the annunciators,

the operators

refer to

ONOP

0208.3,

entitled Annunciator List - Panel

A, Reactor Coolant.

GOP 305,

step

5. 11, directs the operator to establish

and verify OMS

operation.

This is accomplished

by performing OP-041.4,

Overpressure

Mitigation

System.

Section

5. 1

of this

procedure

contains

the

instructions for verifying the

OMS setpoint

and that the

PORVs

open

as indicated

on the appropriate

annunciators.

GOP

305,

Attachment

1, lists the

components

for the cold

shutdown

clearance

tagging

requirements.

The

inspectors

reviewed

selected

completed

procedures

to verify that the

attachment

was being signed

off as complete

as the units were brought to cold shutdown.

GOP

503,

step

5. 16,

has

the

operators

align the

OMS for normal

pressure

operation

per

OP-041.2,

entitled

Pressurizer

Operation.

Section

5.2

of thi s

procedure

contains

the

steps

necessary

to

transfer the

OMS to normal operation.

GOP 305,

step 5.12.1,

specifies that if a pressurizer

bubble is to be

maintained,

the pressurizer

heaters

and/or sprays

are to

be

used to

maintain pressurizer

pressure

in the range of 325 to 375 psig.

There

are

no specific procedures

for performing this evolution.

Tut key

Point

TS,

Section

3. 15,

specifies

that

with

the

RCS

temperature

less

than or equal

to

275F with

RCS pressure

boundary

integrity established,

two

PORVs shall

be operable with a setpoint of

415 psig k 15 psig.

The inspectors verified that the plant installed

system,

along with the

appropriate

procedural

guidance,

exist to

ensure that the

TS requirements

are being followed.

Training and Equipment Modifications,

The inspectors

reviewed

Lesson

Plan

No. 6902009, entitled Pressurizer

and Relief System.

This training is designed

for licensed

operator

candidates,

and contains

a description of the

OMS and the operational

characteristics.

The license

candidates

must

also

perform various

evolutions

under

the

guidance

of

a

licensed

operator.

These

evolutions include testing of the

OMS and aligning the

OMS for normal

and low pressure

operations.

The inspectors

also questioned

selected

RCOs

concerning

the

operation

of the

OMS

and

the

causes

and

prevention of pressure

transients.

All of the operators

interviewed

were knowledgeable

of the

OMS operation

and the necessary

precautions

to be taken to prevent pressure

transients.

19

Turkey Point Administrative Procedure

(AP), 0190.15 entitled Plant

Changes/Modifications

(PC/M), is used to control the review, approval

and

implementation

of PC/Ms.

Section

8. 1 contains

instructions for

the

preparation

of the

PC/M which

includes

preparing

a

safety

evaluation

report

in accordance

with

10 CFR 50.59

to

review

any

unreviewed

safety

questions.

This

procedure

also

requires

the

Startup

Department

to perform functional

tests

after

the

PC/M is

completed.

AP 0190.22,

entitled

Changes,

Tests

and

Experiments,

provides

guidance for preparation

of the safety evaluation

report.

These

admini strative controls provide

an effective method to resolve

any unresolved

safety issue during any

PC/M,

and also to ensure that

functional testing is performed after completion of the modification.

In the

example

given for the modifications of the solenoid

valves,

the

PC/N would be developed

and the post modification testing

would

be

performed

under

the administrative

controls to ensure

that the

equipment

is still in its design

basis.

If maintenance

were to

be

performed

on

the

sol enoi d

val ves,

AP

0190. 28,

enti tled

Post

Maintenance

Testing,

would govern.

Appendix

E to this

procedure

provides specific guidance for post-maintenance

test requirements

for

solenoid valves.

This will also ensure that the design basis of the

system is being maintained.

Drawing

5610-T-D-16B

shows

the

annunciators

provided for the

ONS.

The

OMS

has

an annunciator

that will alert

the

operator

that

RCS

pressure

is approaching

the

OMS setpoint.

This will alarm at

400

.

psig

when the

OMS mode switch is in low pressure

operations

and

the

PORV control switch is in automatic.

The

inspectors

reviewed

the initial test

data

obtained

after the

installation of the

backup

N-2 system

under

PC/M 78-16

and 78-17.

This

data

indicates

that

the

system

design

allows the

PORV to

be

opened

for

at

least

ten

minutes

using

the

backup

N-2

system.

OSP-041.4,

entitled

Overpressure

Mitigating System

Nitrogen

Backup

Leak and Functional Test, is performed periodically to ensure

the N-2

Backup System

can perform its intended function.

Surveillance

The

PORV electronics

and setpoints

are verified each time before the

OMS is aligned for low pressure

operations

per

OP-041.4,

entitled

Overpressure

Mitigation System.

The inspectors

reviewed the stroke

times for the Units 3

and

4

PORVs dating

from May 1984 through

May

1988.

These tests

were

performed in accordance

with the licensee's

Inservice Test (IST) Program.

The

IST requirement

for each

PORV is

to fully stroke

within

15

seconds.

The

stroke -times

reviewed

indicate

an average

opening time as fol.lows:

20

Valve

Stroke time

seconds

PCV"3-456

PCV-3"455C

PCV-4-456

PCV-4-455C

4.25

3.67

2.76

2. 41.

These

average

stroke

times

are

well

below

the

15

second

IST

criterion.

However,

in an

NRC safety evaluation report

(SER) issued

on March 14,

1980,

supporting license

amendment

numbers

55/47 for the

OMS,

a relief valve

opening

time of 2.0

seconds

was

assumed

in

calculating the setpoint overshoot for the mass input case.

The heat

input case

assumed

a relief valve opening time of three

seconds.

The

PORVs,

on the average,

are not meeting the design

basis

stroke times

indicated

in the

SER

~

This discrepancy

is being

reviewed

by the

licensee

and will be tracked

as IFI 250, 251/88-14-03.

e.

Summary.

The inspectors

completed

a review of the

OMS design

and installation

for Turkey Point Units

3 5 4.

TI 2500/19 is closed.

However,

as

described

in paragraph

8.d.

above,

the

adequacy

of

PORV stroke

time

testing

remains

a

followup

item

pending

further

review

and

evaluation.

No violations/deviations

were identified in the areas

inspected.

9.

Plant Events

(93702)

The following plant events

were reviewed to determine facility status

and

the

need for further

followup action.

Plant

parameters

were evaluated

during transient

response.

The significance of the event

was evaluated

along with the

performance

of the

appropriate

safety

systems

and

the

actions

taken

by the

licensee.

The

inspectors

verified that

required

notifications were

made to the

NRC.

Evaluations

were performed relative

to the

need for additional

NRC response

to the event.

Additionally, the

following issues

were

examined,

as

appropriate:

details

regarding

the

cause of the event;

event chronology; safety

system performance;

licensee

compliance

with approved

procedures;

radiological

consequences,

if any;

and

proposed

corrective actions.

The licensee

plans

to

issue

Licensee

Event Reports

(LERs)

on

each

event within 30 days following the date of

occurrence.

On June

15,

1988, at 0835, with Unit 3 at

100% power,

excessive

noise

and

high bearing temperatures

were noted from the main generator air side seal

oil pump.

Unit power

was

reduced

to about

95% and the subject

pump

was

replaced.

The unit was returned to

100% power at 1600

on the

same day.

On June

15,

1988,

at

1654, with both units at

100% power,

the licensee

reported

a significant event in that the

Emergency

Notification System

(ENS)

phone

was

out of service.

The licensee

contacted

the telephone

21

company to effect repairs

and the

ENS phone

was placed

back in service

on

June

16,

1988, at 1520.

Exit Interview

The

inspection

scope

and

findings

were

summarized

during

management

interviews held throughout the reporting period with the Plant Manager

Nuclear

and selected

members of his staff.

An exit meeting

was conducted

on June

29,

1988.

The areas

requiring management

attention

were reviewed.

No proprietary

information

was

provided

to the

inspectors

during

the

reporting period.

Item Number

Descri tion and

Reference'50,

251/88-14-01

250,

251/88-14-02

Violation - Failure to control materials

used in

safety related

systems,

paragraph

5.

IFI - Evaluate

the

root

cause

of using

a

20

ampere

breaker instead of the required

30

ampere

breaker

in the alternate

power supply to the

RPI

system,

paragraph

5.

250,

251/88-14-03

IFI

Evaluate

the

basis

for selecting

maximum

stroke times for the

PORVs,

paragraph

8.

Acronyms and Abbreviations

ADM

AFW

a.m.

ANSI

CFR

FPLTQAR

gpm

ICW

IE

IFI

IST

LCO

LER

MCC

MOV

Administrative