ML17347B435

ll l4 /89

ate

cygne

E1ectri ca 1 Power:

Instrumentation:

Hechanical

Components:

Hechanical

Systems:

Haintenance:

Operations:

Plant Systems:

S.V. Athava le,

0. Hazzoni, Consultant

J. Leivo, Consultant

A. du Bouchet,

Consultant

D. Waters,

Consultant

G. Schnebli,

Region II

J.

Thompson,

K. Poertner,

Region II

Reviewed

By:

ugene

.

Im ro,

se

Team Inspection Section

B

Special

Inspection

Branch,

DRIS,

Il iV

P9

at

s

ed

Special

Inspection

Branc

, DRIS,

Approved By:

'ayn

.

annsng,

se

ate

cygne

8<)12010223

8911202250

PDR

ADOCN; 0500

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TABLE OF

CONTENTS

EXECUTIVE SUMMARY ..............................................

~Pa

e

ES-1

1.0

2.0

I)TRODUCTION ...........................................

TNSPECTION DETAILS

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~

2.1

2.2

2.3

2.4

2.5

2.6

2.7

2.8

2.9

2.10

2.11

Performance

Enhancement

Program

(PEP) ..

Plant

Change

and Modification Program ..

2.2.1 Plant Modification Implementation

Plant Drawings .........................

Design Integration

Review Team .........

Engineering

Design Packages ............

2.5.1

Electrical ......................

2.5.2

Instrumentation

and Control .....

2.5.3

Mechanical

Components ...........

Vendor Surveillance ....................

Systematic

Design Investigation

Program

Select

Systems

Review ..................

2.8.1

Safety Engineering

Group

(SEG) ..

2.8.2

Small Bore Piping ...............

2.8.3

Reactor Protection

System .......

2.8.4

Component

Cooling Water System ..

2.8.5

Electrical Distribution System ..

2.8.6

Plant Configuration .............

Surveillance

Testing ...................

Mainte6ance ............................

Operations .............................

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3

5

6

7

8

9

10

11

12

13

18

19

20

21

22

3.0

3.1

3.2

3.3

3.4

SUMMARY OF LICENSEE STRENGTHS

AND WEAKNESSES ....... "..

W ~

Plant Modifications ....................................

Design Basis

Documents .................................

System Procedures ......................................

Systematic

Design Investigation

Program ................

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22

23

24

4.0

MANAGEMENT EXIT MEETING ................................

24

APPENDIX A - PERSONNEL

CONTACTED ............................

APPENDIX

B - DEFICIENCIES ...................................

APPENDIX C -

PERFORMANCE

ENHANCEMENT PROGRAM INSPECTOR

FOLLOWUP ITEMS

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A-1

B-1

C-1

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Y'1

III

EXECUTIVE SUYitlARY

INSPECTION

REPORT 50-250

and 50-251/89-203

FLORIDA POWER

AND LIGHT COMPANY (FP&L)

TURKEY POINT NUCLEAR POWER PLANT, UNITS 3

AND 4

During the periods of September

11 through 15, September

25 through 29,

and

October

12 and 13,

1989,

a Design Validation Inspection

(DVI) was conducted at

the Turkey Point Nuclear Power Plant, Units 3 and 4,

and the Nuclear Engin-

eering offices at Juno

Beach, Florida.

The purpose of the inspection

was to

assess

the effectiveness

of the actions

implemented

as part of the Turkey Point

Performance

Enhancement

Program

(PEP) Project

4 on configuration control.

To

do so the team performed

a Safety

System Functional

Inspection

(SSFI)

on the

reactor protection,

component cooling and electrical distribution systems.

As a result of the inspection,

the team concluded that the facility had

an

inadequate verification program for the Design Basis

Documents

(DBDs) and

Component

Design

Documents

(CDRs).

The team identified numerous errors within

the

CDRs.

The verification program did not envelope

the external

hazards

design basis

and only addressed

a limited scope of the Reactor Protection

System

(RPS) characteristics.

The existing inaccuracies

in the

CDRs limited

the usefulness

of the

DBDs.

The team identified a number of concerns

regarding the seismic

and anchorage

qualifications for mechanical

components

in the component cooling water

(CCW)

system.

In some cases,

no calculations

were available,

and in other situa-

tions, there were errors in the method of analysis.

The adequacy of the

emergency

powe~ system

was not fully demonstrated

for the case of a battery

end-of-service-life voltage.

The team identified additional errors within the

design output documents

generated

by the architect engineers

in the electrical

and mechanical

areas.

Some

key electrical calculations

were not available.

Finally, some plant operating

procedures

were found in error with respect to

the'alve

lineup configuration.

The inspection

team identified several

strengths.

These

included the

analysis-based

preventive

maintenance

program, which utilized results

from

vibration, oil sampling,

and thermography

equipment monitor ing programs.

The

system level

DBDs were well prepared.

The team also noted significant improve-

ment in the quality of the Plant

Change

and Modification (PC/N) packages

as

a

result of the

PEP actions.

The Systematic

Design Investigation

program

provided

an excellent

means to prioritize and resolve outstanding

technical

issues.

Based

upon the identified design output document

and design interface control

problems,

the team concluded that greater technical oversight is needed

on the

part of the Florida Power and Light Company

(FP&L).

These

included

improperly performed stress

analyses for CCW components

and piping systems,

battery charger specification errors

and failure to revise

a stress

calculation

based

upon increased

loads.

ES-1

II

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Pr

h

1.0

INTRODUCTION

The Nuclear Regulatory

Commission

performed the Turkey Point Design Validation

Inspection

(DVI) to assess

the actions that have

been

implemented

as Project

4

of the Performance

Enhancement

Program

(PEP)

on configuration control.

The

PEP

was developed

by Florida Power and Light (FP8L) in 1984 to address

performance

and programmatic

problems at Turkey Point.

The four main objectives of the

PEP

were the continued

safe

and reliable plant operation,

improved plant material

conditions,

increased

emphasis

on quality performance

and continued responsive-

ness to regulatory requirements.

The

PEP was

composed of the following nine

projects:

site facility upgrade,

operations

enhancement,

procedures,

configu-

ration control program, training, management

action program, licensing program,

quality assurance

and quality control,

and the maintenance

management

system.

The remaining

PEP projects,

one through three

and five through nine,

were

evaluated

by Region II inspections.

The DVI assessed

the implementation of PEP regarding

the configuration control

program.

The program

had been

developed

to provide more detailed

systems

and

controls to describe

and

manage

the facility design

and operation.

The follow-

ing eleven tasks

were associated

with the configuration control project:

Task

1 - Enhancement

of the Plant

Change

and Modification (PC/M) program

by the establishment

of a Plant Review board to screen

proposed

PC/Ms and

issuance

of improved construction work controls.

e.

Task

2 - Evaluation of construction

documentation

to improve the controls

for the

PC/M implementation

and closeout

process.

Task

3 - Generation of an

FPSL startup

manual to formalize the startup

and

turnover process.

Task 4 - Updated drawings to incorporate outstanding

PC/Hs

and developed

criteria to as-built operations critical drawings.

Task

5 - Enhancement

of the control of vendor manual

documentation.

Task

6 - Formation of a Design Integration

Review Team (DIRT) to enhance

the coordination

between engineering disciplines.

Task

7 - Development of a standard

engineering

design

package to ensure

that

PC/Ms met the required

design criteria.

Task 8 - Development of an enhanced

program for vendor surveillance

and

establishment

of a formal review of the Architect Engineer

(A/E) work

products.

Tasks 9, 10,

and ll - Performance of selected

safety

system reviews.

A

Phase

I operability review of ten systems

was performed

by the Safety

Engineering

Group (SEG).

Phase II involved a further review of fourteen

accident mitigating and support

systems

which included:

Reconstitution of design basis

information and verification to

establish

consistency

between

the design basis

and design drawings.

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Performance

of plant walkdown verifications for small bore piping and

tubing isometrics,

piping and instrument diagrams,

operating dia-

grams,

and the breaker list.

Performance

of a

SEG functionality and operability review.

Implementation of an enclosed

configuration management

program.

Implementation of plant hardware modifications

and performance

of an

Auxiliary Feedwater reliability study.

The DYI was performed in the framework of a Safety

System Functional Inspection

(SSFI)

on three plant systems.

The SSFI technique

involves the performance of

a vertical slice review of a system across

the functional areas

of engineering,

surveillance,

maintenance,

and operations

to verify that the attributes of

these functional areas

support the as-built plant configuration.

The systems

reviewed includeg component cooling water, reactor protection

and electrical

distribution.

The inspection evaluated

the adequacy

of the system Design Basis

Documents

(DBDs) by verifying the appropriateness

of the design

bases

and their

incorporation in the inputs to the

DBDs.

The

DBDs were then validated

by

reviewing their consistency with the as-built plant configuration, plant

procedures,

and engineering

calculations.

The inspection

was conducted at both the corporate

nuclear engineering offices

and the Turkey Point facility.

The inspection

dates

were September

11 through

~

15, September

25 through 29, and October

12 and 13,

1989.

This report describes

the activities and findings generated

by this team

inspection.

Some of the findings may result in potential

enforcement

items.

Region II will initiate and execute

any required

enforcement action that

results

from this inspection.

The inspection

team examined

a representative

sample of design

documents,

operating

procedures,

plant drawings, plant components,

and

DBDs to ascertain

the ability of the selected

systems

to perform the design functions.

The

review evaluated

compliance with regulatory requirements

and effectiveness

of

the design control process.

During the course of the inspection,

the team

on

a

sample basis verified the consistency

of the

DBDs with licensing

commitments,

design

and accident

analyses,

system performance criteria,

and system design

documents

and validated the ability of the system to perform the intended

design basis functions by comparing the

DBDs with the as-built plant.

The team

validated the DBDs, assessed

system functionality, and examined

the

PEP

Project

4 corrective actions

by inspecting:

The product of the licensee

Safety Engineering

Group

(SEG) efforts;

The adequacy of the Plant Operating

Diagrams

(PODs);

The ability of mechanical

and electrical

system

and component testing to

ensure

the system or component would perform its required safety function;

System maintenance activities to ensure that material condition will

support reliable system performance;

The operating

procedures

and training;

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The impact of system modifications

on system functionality, adequacy

of

safety evaluations,

and configuration management

practices;

The consistency of the plant configuration with the

DBDs;

CCW system valve alignments for different operating

modes relative to

design

requirements

and

The accuracy of the small bore pipe arid tubing isometric accuracy

and

adequacy

of the analysis;

and

The accuracy of the system piping and instrumentation

diagrams.

Appendix A lists the liceiisee personnel

who were contacted

during the inspec-

tion, as well as those

who were present at the exit meeting.

Appendix

B

documents

the deficiencies that were identified by the inspection

team.

NRC

open item numbers

used within the body of the report correlate with the identi-

fied deficiencies

as described

in Appendix B.

Deficiencies were not written

for the nine inspector followup items identified in Appendix

C because

these

involved

PEP items for which licensee

actions

were outstanding

and for which

inadequate

information was available to inspect.

2.0

INSPECTION DETAILS

2.1

Performance

Enhancement

Pro ram

(PEP

The inspection

team reviewed the eleven task areas

associated

with PEP

Project 4.

This involved an examination of the commitments,

procedural

changes

and program enhancements,

personnel

interviews, records review,

and review of

the

PEP implementation

through the performance of a functional inspection for

three safety-related

systems.

The inspection revealed

several

areas

in which the

PEP actions

remain outstand-

ing and for which inadequate

information was available to inspect.

These

have

been tracked

by the appropriate

licensee

management

systems

and are scheduled

for resolution.

They are identified in Appendix

C as Inspector

Followup

items 89-203-01

through 89-203-09.

The licensee

work has

been prioritized with

respect

to the safety significance of the actions.

With the exception of the

actions that the licensee

was currently pursuing,

the team found that the

facility had satisfactorily

implemented

the

PEP items.

2.2

Plant

Chan

e and Modification Pro ram

The team reviewed the enhanced

process

that had been established

for the Plant

Change

and Modification (PC/H) program

as part of Tasks I and

2 of the

PEP.

The implementing procedures for the Change

Review Team

(CRT) and Plant Review

Board

(PRB) were examined.

The multi-level review process for proposed

PC/Ns

addresses

work prioritization, technical

review, cost estimates,

work scope,

and work description.

The review groups are

composed of plant supervision

and

management for the

CRT and

PRB, respectively.

The review groups

screen

unnec-

essary

work requests

prior to the initiation of engineering

work on the pro-

ject, prioritize the

PC/Ns according to plant needs,

and perform

a technical

review of the proposed

work.

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"'The 'team reviewed

an

FPSL task team assessment

of the large quantity of plant

changes

that had been in progress.

The assessment

was performed to identify

means to ensure that the operations staff had clear

knowledge of the plant

configuration during the modification process.

The task team also evaluated

methods to reduce the

PC/H life cycle and to expedite

PC/H closeout.

The

inspection

team reviewed the associated

plant procedures

and found the facility

had taken appropriate

steps

in response

to the task team recommendations.

The

team also reviewed

a recent

PRB meeting

summary to assess

how the group func-

tioned.

The enhanced

PC/H procedures

and processes

were found satisfactory.

2.2.1

Plant Hodification Im lementation

The team reviewed the following PC/Hs in detail to assess

their implementation:

PC/N 86-101,

EDG-B Auxiliaries Power Supply Relocation

Fuel Oil Transfer

Pump

4P10

and CV-2046B; and

PC/H 86-238; Safety Relief Valve Replacement

Component Cooling Water

System.

PC/H 86-101

addressed

modifications to three

components

in the emergency

diesel

generator

(EDG) fuel oil system.

While the team's

review found that the

facility tested

the fuel oil transfer

pump and control valve CV-2046B following

the modification, the team could not readily identify post-modification test

results for solenoid valve SV-3552B.

Subsequent

review of the diesel generator

surveillance

procedure

indicated that the valve was tested appropriately,

although the facility failed to include documentation with the

PC/N to substan-

tiate the testing.

The team also identified that,

due to

a relabeling of

common plant equipment,

the test documentation for PC/N 86-101

was not consis-

tent with the faci lity 's revised

equipment labeling.

Based

upon the team's

observations

noted in Section 2.9 of this report,

the nomenclature

problem also

existed in the

EDG fuel oil surveillance

procedures.

The post-modification test procedures

were reviewed for PC/H 86-238.

The team

noted that the documentation

was difficult to review and that clarifying

information was required from the responsible test engineer.

The team reviewed the startup testing manual,

issued

by FPSL,

as part of Task

3

of the

PEP.

The manual

was comprised of the Field Startup

Procedures

(FSPs).

The team assessed

the effectiveness

of the startup

manual

through

a review of

post-modification testing.

The

FSPs were found adequate

and the system testing

conformed to the

FSP requirements.

In general,

the team found that PC/Ns written after the

PEP implementation

were

more comprehensive

than the earlier PC/Ns.

For example,

recent

PC/Hs included

startup testing

and pre-operational

requirements

that were lacking in the

earlier

PC/Hs.

With the exception of the test procedure

inconsistencies

within

PC/N 86-101,

the plant modification implementation

phase

was considered

satisfactory.

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'~i'he

team reviewed the piping and instrumentation

diagrams

(P8IDs), training and

education

diagrams

(T-Es),

and the isometric drawings for the

instrument

air,

125V/120VAC vital power,

and

emergency

diesel

generator

(EGG] systems

to

assess

the implementation of PEP Task

4 for drawing updates.

The team performed walkdowns of the Unit 3 and

4

CCW systems with respect

to

the associated

T-E drawings,

P&IDs, and piping isometric drawings.

The team

found that the T-E drawings accurately reflect the installed system.

Minor

inconsistencies

were identified on the piping isometric drawings involving the

location of branch lines

and vent or drain valves.

The facility informed the

team that the current revision of the isometrics

was scheduled for field

verification to update the drawings to an as-built status.

The team found the

system

P&IDs difficult to use

because

of poor legibility.

The licensee currently maintains four separate

sets of P&ID type diagrams:

A set of T-E drawings that are functionally as-built and are maintained

in

the control

room;

A set of P&IDs that depict as-built line sizes;

A set of P&IDs that depict the piping seismic boundaries;

and

A set of P&IDs that depict the piping safety classifications.

The team reviewed the various

P&IDs associated

with PC/M 86-238

and identified

several

discrepancies

in these

drawings.

The line size

P&ID did not specify

correct valve sizes,

and the seismic boundary

P&ID incorrectly identified the

valves

as non-seismic.

However, the safety classification

P&ID did indicate

that the valves were safety-related,

and the licensee

did procure these

1-1/2 inch and I inch valves

as safety-related

and seismically designed.

The

licensee

indicated that these

separate

sets of P&IDs will be consolidated.

Bechtel

had previously revised the original Dravo pipe fabrication drawings to

reflect the installed plant configuration.

Bechtel later reissued

those

drawings

as as-bui lt drawings in response

to

NRC Bulletin 79-14.

This program

was completed in 1984;

however,

those

drawings were neither maintained as-built

nor controlled by the licensee.

For that reason,

as-built piping physical

drawings

prepared

by Teledyne also function as piping stress

isometric draw-

ings.

Teledyne is currently issuing updates

of the Bechtel piping physical

drawings to incorporate

design

changes

that have

been

implemented

since

1984.

These

drawings will be field verified, and as-built versions of the piping

isometrics will be issued.

The licensee

has indicated that these

drawings will

be maintained as-built and controlled in-house.

The team found

some minor errors in the licensee's

instrumentation

and control

( I&C) plant operating

diagrams

(PODs)

as discussed

in Section 2.8.4, but

concluded that these errors would have

no impact on safe operation or design of

the plant.

For example,

the team identified one-line

POD that omitted

a fuse,

which could affect the use of the one-line

by operations.

The team found the

elementary

diagrams to be of good quality and without errors.

However, for

non-POD drawings,

such

as internal wiring and connection

diagrams,

the team

found several

errors

as discussed

in Section 2.5.2.

The wiring diagrams for

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Requirements:

10 CFR 50, Appendix B, Criterion V, states that procedures

shall

be utilized to

govern safety-related activities to ensure that the activities have

been

'" - - 'satisfactorily accomplished.

References:

3/4-0P-030,

Component

5610-T-E-4512)

Sh.

1,

5610-T-E-4512,

Sh. 2,

Cooling Water System,

6/29/89.

CCW System Outside Containment,

Rev. 71.

CCW System Inside Containment,

Rev. 27.

DEFICIENCY 89-203-16:

CCW Pum

and Sur

e Tank Seismic

uglification and

Anchora

e

ec

Discussion:

equipment specification

676428 included the seismic

qualification criteria for the

CCW pumps.

Section 3.2.12 of the specification

stated that the

pumps shall

be designed to resist earthquake

forces in the

horizontal

and vertical directions,

as specified

by the

pump data

sheets.

The

centrifugal

pump data sheet

APCC-532 speci'fied

a horizontal design

acceleration

of 1.0

g and

a vertical design acceleration

of 0.67 g.

FPEL could

not access

the seismic qualification documents for the

CCW pumps.

FP&L addi-

tionally could not access

any seismic criteria for the

CCW surge tank, or any

sei smi c qua 1ification documents.

The equipment

anchorage

should

be checked for the combined effects of piping

thrusts,

deadload

and seismic load.

However,

FPSL could not access

the anchor-

age calculations for the

CCW pumps

and surge tanks.

The team was informed that an essential

calculation

program is planned for the

facility.

lhe licensee will determine

which calculations

are required

and will

verify that the calculations

are retrievable.

This item is unresolved

(89-203-16).

Requirement:

FSAR Appendix

5A designates

the

CCW pumps

and surge tank as Class

I equipment

and requires,

in part, that Class

I equipment is required to be designed to

withstand the appropriate

earthquake

loads applied simultaneously with other

applicable

loads without loss of function.

References:

FSAR Appendix 5A.

equipment specification

67428, Auxiliary Pumps,

Rev. 0, 11/23/66.

centrifugal

pump data

sheet

APCC-532, 1/30/68.

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DEFIClENCY 89-203-17:

Audibilit of the

CCW Stress

Packa

es

Discussion:

Teledyne

reviewed the safety-related,

large-bore piping systems,

equipment,

and supports

associated

with the

CCM system in Units 3 and

4 for

acceptance

to

FSAR criteria.

Bechtel previously reviewed these piping systems

for functionality.

Review of Teledyne calculations

6961C-1

and 6961C-3

revealed that the Teledyne stress

packages

cannot

be audited

as

independent

documents.

The Teledyne calculations

use information which Bechtel originally

prepared,

without clear reference

to 'the originating Bechtel

documentation.

ExampTes of such unreferenced

information include equipment nozzle thermal

displacements

and valve weights

and offsets.

The Teledyne stress

packages

do not appear to incorporate, either directly or

by reference,

the Bechtel information required to make these stress

packages

audi table.

This item an Inspector

Follow-up item (89-203-17).

Requirement:

Section 4.1 of ANSI N45.2.11 requires that design activities be documented

in

sufficient detail to permit verification and audit.

References:

Teledyne calculation 6961C-l, Analysis of Stress

Problem 025, Unit 4 for

Replacement

of

CCW Heat Exchangers,

Rev. 3, 11/30/88.

Teledyne calculation 6961C-3, Analysis of Stress

Problem 038, Unit 4, for

Replacement

of

CCW Heat Exchangers,

Rev. 5, 10/28/88.

DEFICIENCY 89-203-18:

CCW Pi

e

Su

ort Calculations

Discussion:

The team reviewed the calculations for approximately

twenty four

pipe supports

which were documented

in the following Teledyne stress

packages:

Teledyne calculation TR-5322-93,

USNRC

ISE Bulletin 79-14 Analysis,

Turkey Point Unit 4 Nuclear

Power Plant,

Component

Cooling Mater System

(Outside Containment)/Stress

Problem CCW-14, Revision 1, dated

November 21, 1984;

Teledyne calculation 6961C-1, Analysis of Stress

Problem 025 Unit 4,

Turkey Point, for Replacement

of CCM Heat Exchangers,

Revision 3, dated

November 30, 1988;

and

Teledyne calculation 6961C-3, Analysis of Stress

Problem

038 Unit 4,

Turkey Point, for Replacement

of CCM, Revision 5, dated October 28,

1988.

The team

compared

these calculations

against the applicable Teledyne engineer-

ing procedures

and identified the following calculational

and procedural

deficiencies:

Teledyne re-qualified

a number of stanchion

supports

to resist uplift.

However, the Teledyne baseplate

procedure

does not appear applicable to

B-7

~

~ I

t

<4 s

I

~

.

~ -the qualification of these

supports.

Pipe support.4-ACH-267 is an example

of such

a stanchion

support.

0

Some anchor bolt tens>on

and shear

loads,

such

as for support SR-703,

were

not computed in accordance

with the Teledyne procedure.

For example,

baseplate

edge distance amplification factors were not applied to compute

bolt tension

loads,

and the shear

loads were distributed to all, rather

t5an half, of the anchor bolts.

The allowable bolt tension

used to qualify pipe support

4-ACH-211 exceeded

the bolt tension allowed by the Teledyne design guide.

Bending stresses

in single-angle

supplementary

steel

were not correctly

computed.

Examples

included the supplementary

steel for pipe

supports

4-ACH-14 and 4-ACH-46.

Some supplementary

steel

was checked

using

assumed

cross-sectional

dimen-

sions that were not field verified.

Examples

included the supplementary

steel for pipe supports

4-ACH-190 and 4-ACH-191.

Spring hanger

4-ACH-207 tops

and bottoms out, but was accepted

as-is

without ana lysis.

The Teledyne stress

packages

indicated that

ZPA and seismic inertia loads

should

be combined absolutely,

however,

these

values

were actually

calculated

by the

SRSS

method within the stress

package.

The AISC web crippling check was not performed to determine if beam

stiffeners are required.

This item is unresolved

(89-203-18).

Requirement:

10 CFR 50, Appendix B, Criterion V, states that safety-related activities shall

be performed in accordance

with the appropriate instructions.

References:

Teledyne technical report TR-5322-1, Project Procedures

and Criteria/VSNRC

ISE Bulletin 79-14 Analysis/Turkey Point Units 3 and 4, Rev. 1,

January

29,

1982.

Teledyne engineering

procedure

EP-2-024, Amplification method for Baseplate

Prying,

Rev.

1, April 30,

1982.

Teledyne project guide 5322/5875,

Anchor Bolt Allowables - FSAR Piping System

Support Analysis - Turkey Point Units

3 and 4, Rev. 1, January

23,

1984.

B-8

0

ft

DEFICIENCY 89-203-19:

Small-Bore Pi

e

uglification

Discussion:

Bechtel

walkdown package

CCM-3-III-1 and backup Bechtel calcula-

tion C-499-167 were reviewed.

The team assessed

the qualification of the

branch lines to the governing criteria of Bechtel specification

5177-PS-21,

Project Implementation of User's

Manual tl-18 for Routing and Supporting

2 Inch

and Under Piping for tlodification to Turkey Point Units 3 and 4.

The calcula-

tion accepted

two branch lines with frequencies

of 22-24

Hz without requiring

tieback supports to the piping run.

The tieback supports

are required

by the

Bechtel specification for branch 'lines with fundamental

frequencies

less

than

33 Hz.

This item is an Inspector Follow-up item (89-203-19).

Requirement:

10 CFR 50, Appendix B, Criterion V, states

that activities affecting quality

shall

be accomplished

in accordance

with the governing procedural

requirements.

Bechtel specification

5177-PS-21

requires that branch lines with a fundamental

frequency

less

than

33

Hz be supported off the run pipe with tieback supports.

References:

Bechtel

walkdown package CCW-3-III-l, CCM Instrumentation at Emergency

Cooler,

Rev. 0, 3/31/87.

Bechtel calculation C-499-167,

Review and Evaluation of Walkdown Package

~

CCM-3-III-1, Rev. 1, 8/25/89.

Bechtel specification

5177-PS-21,

Project Implementation of Users

Manual N-18

for Routing and Supporting

2 inch and Under Piping for Modification to

Turkey Points Unit 3 and 4,

Rev. 2, 2/7/89.

DEFICIENCY 89-203-20:

Com onent Desi

n

Re uirements

Discussion:

The

CDRs for the

RPS,

CCW, and electrical distribution systems

were reviewed.

The

CDRs were found to contain erroneous

and unnecessary

information, as follows:

The

CCW chemical mixing pot was inferred to have

been replaced

when it had

only been

moved to a different location on the system.

The

CCW pump start

sequence

description

was incomplete.

The

CCW containment isolation valve stroke times were not consistent with

the

TS requirements.

The instrument voltage tolerances

were improper ly specified.

The methodology for cable short circuit calculations

was inconsistent with

actual practice.

The battery discharge profile was ambiguously defined.

B-9

~

I

~

0

'These

con'cerns

led to the conclusion that the

CDR had not been appropriately

verified.

The licensee

has

agreed to perform some additional verification of

the

CDR information.

The licensee

also issued

a directive which restricted

the

use of CDR information in the design

process.

This item is unresolved

(89-203-20).

Requirement:

10 CFR 50, Appendix B, Criterion III, states

that the design basis

requirements

shall

be translated

in a controlled manner into design

documents.

References:

Component

Design Requirements

Document 5610-030-DB-002,

Component Cooling

Water System,

Rev. 0.

Component

Design Requirements

Document,

5610-023-DB-002,

Emergency

Power

System,

Rev. 0;

Component

Design Requirements

Document,

5610-003-DB-002, Vital AC/DC System,

Rev. 0.

DEFICIENCY (89-203-21):

Plant

0 eratin

Dia ram Errors

Discussion:

The team reviewed the

POD logic diagram for the component cooling

water

pumps

and determined that the

CCW pump

3B breaker close logic was in

error.

The diagram incorrectly indicated that

a loss-of-offsite-power

load

shedding

signal would close the breaker

and start the

pump.

The correct logic

would open the breaker to clear the loads

on the bus

when

a load shed signal

were present.

The team determined that the elementary

diagrams

implemented

the

cor rect logic.

The licensee initiated DCR-TPI-89-187 to correct the error on

the logic diagram.

The diagram incorrectly depicted the start/stop

selector switch action for all

CCW pumps except

pump 3B.

The logic diagram

shows the switches

as maintained

in the stop position, which does not agree with the elementary

diagram.

These

switches

are spring returned to middle from start

and stop,

and pull-to-lock in

the stop position.

The team also identified an error on the

CCW T-E diagram where the drawing

incorrectly shows

a functional connection

between

the

CCW pumps

and the

pump discharge

pressure

channel

PI-*-640C.

The pressure

channel is a local

gauge with no electrical output,

and the

pump start function is provided by

PC-*-611 which is correctly shown elsewhere

on the drawing.

During a field inspection of miscellaneous

relay rack 4(R46, the team identi-

fied a drawing discrepancy

regarding the termination of PC-611X relay contacts

wired out for the

CCW low pressure

alarm.

The field inspection

indicated that

contacts

3 and

7 appear to be wired out, whereas

the instrument

loop diagram

shows contacts

4 and

8 wired out.

As wired in the field, the contacts will

B-10

~ ~

I

operate

opposite to the contacts

as

shown

on the drawing.

The correct configu-

ration depends

on the external

annunciator circuit, which the team did not

review.

The licensee

agreed to investigate this discrepancy.

The errors in the

POD drawings

do not represent

hardware errors

and are consid-

ered comparatively minor, although they could mislead

an inexperienced

user.

The

POD errors cited herein appear to be isolated errors not typical of other

PODs examined

by the team.

The licensee

agreed to investigate

and correct the remaining errors

and

discrepancies.

This item is an Inspector Follow-up item (89-203-21).

Requirement:

10 CFR 50, Appendix B, Criterion V, states

that appropriate

drawings shall

be

utilized to accomplish activities affecting quality.

References:

FP8L Drawing 5610-T-L1, Sh.

24E, Logic Diagram,

Primary System

Component

Cooling Water Pump,

Rev. 0, April 9, 1987.

FP8 L Drawing 5613-E-25,

Sh.

2A, Elementary

Diagram,

Reactor Auxiliaries

Component Cooling Water

Pump

3A Breaker

3AA12, Rev. 0, June

19,

1989.

FP8L Drawing 5613-E-25,

Sh.

2B, Elementary

Diagram,

Reactor Auxiliaries

Component Cooling Water

Pump

3B Breaker

3AB13, Rev. 0, July 24,

1989.

FPSL Drawing 5613-E-25,

Sh.

2C, Elementary

Diagram, Reactor Auxiliaries

Component

Cooling Water

Pump

3C Breaker

3AA17, Rev. 0, June 26,

1989.

FP8L Drawing 5610-T-L1, Sh.

24D, Logic Diagram,

Primary System

Pumps,

Component

Cooling Pumps,

Rev. 4, April 10,

1987.

FP8L Drawing 5610-T-E-4512,

Sh.

1, Component Cooling System Outside

Containment,

Rev. 71, June

13, 1989.

FP8L Drawing 5610-M-401C-96,

Sh.

101, Instrument

Loop Diagram,

Rack 46

Rear - Miscellaneous

Relay Racks,

Rev. 6, August 14, 1989.

DEFICIENCY 89-203-22:

Assurance

of Un rounded

DC Control

Power

Discussion:

An automatic start signal

on low CCW header

pressure

(60 psig) is

provided to the standby

CCW pump.

No credit is assumed for this automatic

start function for safe

shutdown or accident mitigation functions,

and the

automatic start is overridden during

a loss of offsite power to preclude

EDG

over load.

The low-pressure start circuits for each

CCW pump use

a

common

dc-powered control relay located in the miscellaneous

relay rack.

This device

presented

a potential

common

mode vulnerability to all three

CCW pumps.

The

team performed

a limited field inspection of the wiring of this relay and

determined that the contact output wiring for the two trains of pump controls

and the non-safety,

low-pressure

alarm circuit was bundled together.

The team

also identified

a discrepancy

between

the as-built wiring of the relay alarm

contact

and the instrument

loop diagram.

This discrepancy

is discussed

sepa-

rately in Deficiency 89-203-21.

I

'I

ra

y. Il

p I[

Ql

t

Y

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The team's

review identified the following:

The miscellaneous

relay racks were not originally intended to contain

safety-related

circuits.

The team reviewed the relay rack seismic docu-

'entation that demonstrated

the seismic qualification of the rack and the

relay.

The configuration appeared

vulnerable to a single failure which could

disable all three

CCW pump control circuits by blowing fuses

in all three

control circuits.

The licensee

responded

that credit is being taken for

ungrounded

DC control circuits,

a single ground

on either

DC system would

not blow the fuses,

and such ground faults are detectable

and monitored

continuously.

However, the off-normal procedure for clearing grounds

does

not require

immediate action when'a

ground is detected

and provides

no

time limits on operating with a single ground

on the system.

The licensee

developed

a preliminary

FHEA for this circuit to address

actuation failure

of the relay, spurious actuation of the relay,

and contact welding or

short circuits.

The

FLEA took credit for detecting, isolating,

and

repairing the first ground in a timely manner.

Although unlikely, multi-

ple grounds

would need to be assumed if there are

no time limits on

removing grounds while the plant is at power.

The team found that external field cabling from the miscellaneous

relay

racks to the

CCW pumps

and to non-lE had

no separation

of internal wiring.

The team determined that the

FNEA demonstrated

that

a single failure

postulated for the internal rack wiring would not disable both trains.

~

The licensee's

design basis for the single failure criterion is provided in

guality Instruction JPE-gI 2.3.

That instruction specifies that both active

and passive electrical failures must be considered,

and the effect of each

single failure may be analyzed separately.

Accordingly, the ability of the

system to perform its safety function must not be precluded

by a single active

or passive failure.

While the team concluded that design basis

and

FSAR separation criteria have

been

met for the circuits in question,

the team also determined that assurance

must be provided that grounds

on the

DC system will not exist indefinitely.

Presently,

no such assurance

appears

to be provided in plant procedures.

This

is the remaining Inspector Follow-up item (89-203-22).

References:

FPEL Drawing 5613-E-25,

Sh.

2A, Elementary

Diagram,

Reactor Auxiliaries

Component Cooling Water

Pump

3A Breaker

3AA12, Rev. 0, June

19,

1989.

FPKL Drawing 5613-E-25,

Sh.

2B, Elementary

Diagram,

Reactor Auxiliaries

Component

Cooling Water

Pump

3B Breaker

3AB13, Rev. 0, July 24,

1989.

FP&L Drawing 5613-E-25,

Sh.

2C, Elementary

Diagram, Reactor Auxiliaries

Component Cooling Water

Pump

3C Breaker 3AA17, Rev. 0, June 26, 1989.

FPSL Drawing 5610-M-401C-96,

Sh.

101, Instrument

Loop Diagram,

Rack 46

Rear - Miscellaneous

Relay Racks,

Rev. 6, August 14,

1989.

B-12

v

I

T

~ 1

~

~ ~

r

FPKL Off-Normal Operating

Procedure

9608.1,

125

V DC System - Location of

Grounds,

Harch 24,

1989.

FSAR Section 8.2.2,

Figure 8.2-12.

AEC General

Design Criterion

(GDC) 1, Quality Standards,

July 10,

1967

Draft.

FPEL Quality Instruction JPE-QI 2.3, Classification of Structures,

Systems,

and Components,

November 13, 1985.

FPSL Drawing 5177-E-302,

Sh. 25, Electrical Installation - Raceway

Notes,

Symbols,

and Details,

Rev. 3.

FPSL

DBD 5610-000-DB-001,

Selected

Licensing Issues,

Section VI, Electrical

Separation Criteria, para. 2.2.3,

Rev. 0.

Letter to FPKL, FPL-M-441, Tur key Point Units

3 and 4

Instrumentation

and Control Relays,

September

12,

1967.

Telecon note,

Hehrnaz

Djahanshahi

(Bechtel

PBG) to Jim Hiller (Westinghouse

NSID), Westinghouse

BF Relays,

September

27, 1989.

Telecopy

memoranaum

and attached certifications,

Jim Hiller (Westinghouse

NSID)

to Hehrnaz

D'ahanshahi

(Bechtel

PBG), Certificates of Qualification for

BF Relays,

September

27,

1989.

Equipment Specification

G677033,

Relay Racks,

Rev. 1,

October 4, 1968.

memorandum regarding qualification of the Hiscellaneous

Relay

Racks,

authored

by P. J. Horris, September

28,

1989.

FP8L Procedure

3-0SP-30.5,

CCW Pumps

Low Header

Pressure

Start Test,

November 19,

1987.

FPSL Off-Normal Operating

Procedure

3-0NOP-30,

Loss of Component Cooling

Water,

Hay 16, 1989.

Bechtel Letter SFB-3608,

Conduit Separation Criteria (Turkey Point),

January

29,

1987.

DEFICIENCY 89-203-23:

Acce tabi lit of the Hinimum Batter

Terminal Volta e

Discussion:

The minimum end-of-service-life battery terminal voltage is 105

Volts Direct Current

(VDC).

There is no evidence that the terminal voltage is

adequate

to power all safety-related

devices.

The licensee

has

performed

individual voltage drop calculations for load addition or modification.

The licensee

has not performed

a bounding calculation to show that all devices

located remotely from the battery

bus will be able to operate, successfully.

While some tests

have

been performed, certain

components

were bypassed

during

the testing

and were therefore

not verified to operate at the low battery

terminal voltage.

Adequate

assurance

does not exist that the combination of the minimum battery

terminal voltage

and system voltage drop considerations will yield sufficient

equipment voltages to maintain equipment functionality.

This item is unresolved

(89-203-23).

B-13

~

Pg

r ~

, c4'

~

'.-Requirement:

General

Design Criteria

(GDC)

17 and Turkey Point

GDC 39 state that alternate

power systems

shall

be designed with adequate

capacity to permit proper

,";- functioning of the engineered

safety features.

References:

FSAR, pages

8.1-2 and 8.1-3.

DBD 5610-003-DB-002,

page

13, paragraph

3.2.6.

Calculation 5177-272-E06,

125

VDC Circuits - Permissible

Lengths,

Rev. 0,

10/23/86.

Calculation 5177-272-E08,

125

VDC Modified Circuits Voltage Drop for

Battery 3A, Rev. 0, 10/25/86.

B-14

~l

e'

0

APPENDIX

C

PERFORMANCE

ENHANCEMENT PROGRAM INSPECTOR

FOLLOWUP ITEMS

Some licensee

PEP actions

remain outstanding

and insufficient information was

available for inspection.

Further

NRC followup will be necessary

to verify

satisfactory

closure of the following FP&L actions:

Resolution of the

SEG open items from the

Phase

I and

Phase II reviews

(89-203-01);

Issuance

of the

SEG reports for the plant radiation monitoring, area

radiation monitoring,

and normal containment

cooling systems

(89-203-02);

Implementation of the residual

heat

removal valve replacement

and

emergency

diesel

generator air start modifications (89-203-03);

Resolution of the select

systems

punchlist (89-203-04);

Resolution of the

DBD verification open items (89-203-05);

Completion of the breaker list verification and nonconformance

resolution

(89-203-06);

e

Completion of the Unit 3 piping and instrumentation

diagram walkdowns

inside containment

and associated

nonconformance

resolution

(89-203-07);

Completion of the small bore pipe support upgrade

program (89-203-08);

and

Review of the control of vendor manuals

(89-203-09).

The above

items are considered

Inspector

Followup Items.

'4

op

~

--The team-reviewed

the Limiting Condition for Operations

(LCO) reduction efforts

on the

CCW system.

The efforts focused

on the

CCW heat exchangers

since they

had contributed to 84% of the

CCW system

LCO hours.

The licensee efforts

included the installation of a continuous

mechanical

tube cleaning

system,

....; .installation of replacement

Unit 4 heat exchangers,

and testing of a

new Unit 3

continuous

chemical injection cleaning

system.

As a result of the modifica-

tions, the licensee

has achieved

a noticeable reduction in the Unit 4

LCO

hours..

The team considered

the maintenance

program to be satisfactorily

implemented

with an aggressive

analysis-based

preventive maintenance effort.

2.11

0 erations

The team reviewed the licensee's

training

programs

as well as the incorporation

of PC/Ms into the operator training cycle.

All PC/Ms are reviewed

by the

training department

and are

summarized

in a computer database.

A training

report or a training summary

was generated

based

upon the information in the

database.

The training report was sent to the appropriate

personnel.

The

licensee

stated that significant training summaries

are written as training

briefs.

The training briefs are sent from the training department

to the

respective

engineering

departments

for departmental

seminars.

Training summa-

ries require the cognizant plant employees

to attend training lectures at the

training department.

The team reviewed the operations

department

oversight of industry and

regulatory

communications

and found the system satisfactory.

The licensee

had

the necessary

programs

in place to keep operational

personnel

apprised of the

appropriate

technical

information.

3.0

SUMMARY OF LICENSEE

STRENGTHS

AND WEAKNESSES

The following strengths

and weaknesses

represent

the more significant items

that were identified by the team during the DVI.

3.1

Plant Modifications

~Stree

the

The team found that the more recent

PC/Ms sampled during the inspection

were of

generally high quality and represented

a great

improvement over earlier

PC/Ms.

The team was impressed

by two recent instrumentation

and control modifications

which provided thorough

and explicit safety evaluations,

clearly stated

the

bases for the modifications,

and provided explicit design requirements.

The

team found the electrical

and mechanical

PC/M packages

well assembled

from an

administrative standpoint.

Weaknesses

The calculations

and specifications

associated

with the

PC/Ms that had been

generated

by the A/Es contained technical errors.

Contributing causes for the

problems

were the lack of sufficient FPIIL technical oversight of the A/E design

process

and the lack of

FPImL maintained

design specifications to ensure

consis-

tent design

approaches

among the A/Es.

22

~ i

I~

~

t

3.2.- Desi

n Basis

Documents

~Stren the

In general,

the team

was impressed

by the depth, breadth,

and quality of the

system

DBDs and found them to be useful

and well researched

documents for

capturing the design intent without being unduly encumbered

by configuration

detailZ.

The SLI DBDs were also

deemed well researched

and useful.

Weaknesses

The engineering

personnel

did not appear to regard the

DBD as

a design tool.

The fact that the

DBD was produced

by non-licensee

personnel

appeared

to have

contributed to the lack of FPSL acceptance.

Furthermore,

the licensee's

intention to revise the

DBD only once

a year would tend to limit its usefulness

for modification packages.

The breadth of the

DBD verification was insufficient in, some cases.

The

was only verified with respect to a limited set of attributes,

and the alter-

nate verification approaches

were not well specified.

The preparation of a

DBD

verification attribute matrix captured

the alternate verification methods in a

concise

manner.

The

DBD system level documents

were

complemented

by the information contained

within the

CDR, although

numerous errors were found within the

In some

cases

calculations to support the

CDR were unavailable

in the seismic

and

electrical areas.

These

shortcomings

could have

been

avoided if the licensee

had expanded its verification effort to include the

The licensee's

development of the

CDR was considered

a significant area of

weakness,

since the

CDR was originally represented

by the licensee

as

an

approved

and controlled document containing information suitable for use in the

design

process

by utility engineering

personnel

and outside contractors.

3.3

S stem Procedures

Weaknesses

The team found two test procedures

which improperly identified system valves

due to a relabeling of the plant components.

While the correct valves were

manipulated

during the test, test personnel

did not utilize the procedure

update

process

when necessary

to correct procedural

deficiencies prior to test

performance.

Errors were also found within the

CCW system operating

procedure

involving inaccurate

valve lineups.

These

involved situations

where value

positions

were incorrectly specified that could result in a loss of inventory.

The battery surveillance

procedure

was found to test the battery in an improper

sequence

as the service test could be performed in a condition that was not

representative

of the as-found battery condition.

The off-normal operating

procedures for loss of

CCW flow could not be fully implemented

due to

improperly staged

emergency

equipment.

23

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y

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y

~

3.4

S stematic

Desi

n Investi ation

~Stren th

The'syst'e'mat'ic

design investigation

program provided

a risk-based

methodology

to prioritize the technical

concerns resulting from the

DBD reconstitution

program.

The problems

were prioritized based

upon the safety significance of

the cohcern.

The concerns

were evaluated

on

a time frame commensurate

with the

relative risk of the item.

A concern of high risk would be evaluated

immedi-

ately,

and actions including plant shutdown

could be triggered.

Thus, the

attention of engineering

and operations

was appropriately

focused

on issues

that involved inherent high plant risk.

The proceduralization

and expansion of

the program to issues

beyond the scope of the select

system review is consid-

ered

a strength.

4.0

MANAGEMENT EXIT MEETING

The inspection

team conducted

an exit meeting

on October 13,

1989.

The

licensee

personnel

identified in Appendix

A attended

the meeting.

The team

members

presented

the inspection findings contained within this report.

Other

NRC personnel

at the meeting included:

Mr. Eugene

Imbro and

Mr. Gordon Edison from NRR,

as well as Mr. Ross Butcher, Mr. Marvin Sinkule,

and Mr. Frank Jape

from Region II.

e

24

~

IJ

APPENDIX A

PERSONNEL

CONTACTED

R.

G

J.

J

E.

p.

C

R.

J.

J

J.

W.

D

S.

L.

A.

R.

J.

R.

J.

F.

F

R.

R.

J

K.

S

J

J

J

J

V.

J.

D.

J.

J

D.

G

M.

A.

p.

M

J.

L

H.

Acosta

Adams

Anderson

Arias, Jr.

Arnold

Banaszak

Bible

Bleeker

Burford

Burke

Busch

Chancy

Cornell

Craig

Dunston

Earl

Ellis

Englmeier

Evans

Farzam

F lugger

Gallagher

Gianfrencesco

Goldberg

Greene

Gula

Hartzog

Hays

Hosmer

Hutchinson

Kaminskas

Kenney

Koenicke

Kovari k

Krumins

Lanier

Madden

Manyard

Martinez

Morris

Musrock

Osborne

Pabst

Paduano

Acting Vice President,

Nuclear Energy

Engineer,

Design Basis Reconstitution

guality Assurance

Supervisor

Assistant to the Plant Manager

Manager,

Safeguards

Systems - Westinghouse

Supervisor Engineering

Supervisor,

I&C Engineering

Electrical

Lead Engineer

Project Engineer

Contractor - Impell

Supervisor,

I&C

Electrical Engineer

Director, Nuclear Licensing

Lead Hechanical

Engineer

Manager,

Nuclear

Engineering Support Services

Engineer - Bechtel

equality Control Supervisor

CCW System Engineer

Director of guality Assurance

Site Document Control Supervisor

Senior Civil Engineer -

Bechtel

SDI Program Manager

Engineer - Bechtel

Assistant Maintenance

Superintendent

Executive Vice President

Civil Engineering Supervisor

Senior Mechanical

Engineer

Licensing Engineer

Nuclear Engineering

Services

Manager

Director, Nuclear Engineering

Plant Support Manager

Technical

Department Supervisor

Mechanical

Department

Production Supervisor

Lead Engineer - Civil

Supervisor,

I&C

Mechanical

Engineering

Hanager

Configuration Control Engineer

Principal Engineer,

Nuclear Licensing

Electrical Supervisor - Bechtel

Lead Engineer,

Testing

Manager of I&C Systems

Licensing - Westinghouse

l&C Engineer

Lead Instrumentation

and Control Engineer

Design Reconstitution

Manager

Manager of Nuclear Engineering Technical

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'E'. Pearce

C. Pell

R. Pernisi

D. Powell

R., Proctor

R.V. Rajan

R. Diaz-Robainas

R. Rope

J. Santangelo

F. Schiffley

A. Schi ldkraut

B. Sharp

J.

Sharp

W. Skelley

D. Smith

K. Strickland

J. Strong

F. Varona

R. Wade

M. Wayland

J.

Webb

Supervisor

Engineering

Assistant to the Executive Vice President

Engineer - Bechtel

Regulatory

and Compliance Supervisor

Chief Maintenance Electrician

Senior Maintenance

Engineer

ISC Engineer

Supervisor Configuration Control

Engineer - Teledyne

Nuclear Engineer

Electrical Supervisor - Ebasco

Assistant Superintendent

Planned

Maintenance

Root Cause Analysis Supervisor

Nuclear Engineering

Manager

Manager, Electrical

and

ISC Engineering

Design Basis Reconstitution

Mechanical

Department Supervisor

Supervisor,

Engineering

Engineering Project Supervisor,

Plant Support

Maintenance

Superintendent

Assistant Superintendent,

Planning

and Scheduling

Designates

licensee

personnel

who attended

the exit meeting

on

October 13,

1989.

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APPENDIX

B

DEFICIENCIES

This appendix

documents

the deficiencies that were identified by the inspection

team.

NRC open item numbers correlate with the deficiencies

discussed

in the

body of the report.

In addition to those

presented

in this Appendix,

Appendix

C identifies nine open items (89-203-01

through 89-203-09) for which

deficiencies

were not written because

they represent

PEP items for which the

licensee

actions

were outstanding

and inadequate

information was avai lable for

inspection.

DEFICIENCY 89-203-10:

CCW Heat Exchan

er

Fundamental

Fre uenc

Discussion:

FPSL engineering

package

PC/H 88-263

was written for replacement

of the Unit 4 heat exchangers.

FP&L prepared

purchase

order C88658 90314 to

procure the replacement

heat exchangers.

Appendix

C of the purchase

order

required that the replacement

heat exchangers

be qualified by the response

spectrum

approach for the

SSE depicted in Figure

1 of that Appendix.

Bechtel calculation C-SJ-183-02,

CCW Heat Exchanger

Support Pedestal

Load

Evaluation,

included

an evaluation of the heat exchanger

fundamental

frequency.

The Bechtel calculation

computed

a fundamental

frequency greater

than

33

Hz for

the heat exchangers,

and concluded that the heat exchangers

were rigid.

The

calculation therefore

used the Zero Period Acceleration

(ZPA) values of the

SSE

spectrum to compute the seismic reactions of the heat exchangers.

However, the

Bechtel calculation did not consider the transverse flexibilityof the concrete

pedestals

supporting the heat exchangers.

If the heat exchanger

and the

supporting

concrete

pedestals

were analyzed

as

a single mathematical

model, the

fundamental

frequency of the heat exchanger

along its longitudinal axis drops

to about

10 Hz.

This would increase

the magnitudes of the seismic

loads for

which the heat exchanger

must be qualified.

The replacement

heat exchangers

were qualified for piping thrust,

deadloads

and

seismic

loads

by Target Technology Ltd.

Bechtel

recommended

that Target

compute the heat exchanger

fundamental

frequency

and use the

ZPA loads both to

qualify the heat exchangers

and to compute the heat exchanger

support reac-

tions.

Like Bechtel, Target

computed the heat exchanger

fundamental

frequency

without considering

the transverse flexibilityof the concrete

pedestals

and

concluded that the heat exchanger

was rigid.

As a consequence,

Target quali-

fied the heat exchanger with respect

to the

ZPA seismic loads.

Since the heat

exchanger

and concrete

pedestal

configuration is flexible, the Target stress

report does not adequately

qualify the replacement

heat exchangers

for the

governing seismic loads.

This item remains

unresolved

(89-203-10).

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-Requirements:

FSAR Appendix 5A designates

the

CCW heat exchangers

as Class I equipment

and

requires,

in part, that Class

I equipment

be designed

to withstand the

appropriate

earthquake

loads applied simultaneously with other applicable

loads

without loss of function.

References:

FSAR Appendix 5A.

PC/N 88-263,

Component Cooling Mater Heat Exchanger

Replacement.

Purchase

Order C88658 90314, 7/25/88.

Bechtel calculation C-SJ-183-02,

CCW Heat Exchanger

Support Pedestal

Load

Evaluation,

Rev. 5, 1/18/89.

guality Instruction JPN-(1-3.1,

Rev.

15, 7/89.

Bechtel Specification

5177-H900, gualification of Seismic Category

I

Equipment, 4/87.

Target Technology. Report, Stress

Analysis of

CCW Heat .Exchanger

Replacement/

Turkey Point,

Rev.

11, 12/13/88.

DEFICIENCY 89-203-11:

Shell-Side

Nozzle Loads for Re lacement

CCM

Heat

xc an ers

Discussion:

FPSL engineering

package

PC/M 88-263

was written for replacement

of

the Unit 4 heat exchangers.

Teledyne calculation 6961C-l, Analysis of Stress

Problem 025 Unit 4, Turkey Point, for Replacement

of CCM Heat Exchangers

included the qualification of the

CCW piping attached

to the

CCW heat exchanger

shell-side

nozzles.

In order to reduce the shell-side

nozzle loads,

Teledyne

input circumferential

and longitudinal rotational spring constants

at the

pipe-nozzle interfaces

instead of modeling these

interfaces

as rigid anchors.

However, Teledyne did not input a translational

spring in the global

2-direction to account for the transverse flexibilityof the concrete

pedestals

supporting the heat exchanger.

The addition of this translational

spring

constant

to the piping mathematical

model

may change

the frequency

response

of

the attached

piping,

and

may affect the magnitudes of the piping stresses

and

the shell-side

CCW heat exchanger

nozzle loads.

This item is an Inspection Follow-up item (89-203-11).

Requirement:

FSAR Appendix 5A designates

the

CCW heat exchangers

as Class

I equipment

and

requires,

in part, that Class

I equipment

be designed

to withstand the appro-

priate earthquake

loads applied simultaneously with other applicable

loads

without loss of function.

References:

FSAR Appendix 5A.

PC/N 88-263,

Component Cooling Water Heat Exchanger

Replacement.

Teledyne calculation 6961C-l, Analysis of Stress

Problem 025, Unit 4,

Turkey Point,

Replacement

of

CCW Heat Exchangers,

Rev. 3, 11/30/88.

guality Instruction JPN-(1-3.1,

Rev.

15, 7/89.

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DEFICIENCY 89-203-12:

uglification of the Concrete

Pedestals

for the

Re

acement

C

eat

xc an ers

Discussion:

FPEL engineering

package

PC/M 88-263

was written for replacement

of the Unit 4 heat exchangers.

Bechtel calculation C-SJ-183-02,

CCW Heat

Exchanger

Support Pedestal

Load Evaluation,

included

a check of the heat

exchanger

concrete

pedestals

for the replacement

heat exchanger

seismic reac-

tions.

The Bechtel calculation

computed the heat exchanger

seismic reactions

using.-ZPA loads,

which implicitly assumed

the concrete

pedestals

behave

as

rigid structures.

However, Bechtel also did not access

the concrete

pedestal

detail drawing.

Without the civil drawing for the concrete

pedestal,

the load

transfer

between

the heat exchanger

and the top of the pedestal

through the

pedestal

anchor bolts, the structural

capacity of the pedestal itself, and the

load transfer

between

the base of the pedestal

and the building concrete

slab,

cannot adequately

be checked.

Calculation C-SJ-183-02

therefore

contained

several

undocumented

engineering

judgements.

This item is an Inspector Follow-up item (89-203-12).

Requirement:

FSAR Appendix 5A designates

the

CCW heat exchangers

as Class

I equipment

and

requires,

in part, that Class

I equipment

be designed

to withstand the appro-

priate earthquake

loads applied simultaneously with other applicable

loads

without loss of function.

ANSI N45.2.11 requires that design inputs, including component physical inter-

faces,

shall

be defined

as necessary

to permit the design activity to be

performed in the correct manner.

References:

FSAR Appendix 5A.

PC/M 88-263,

Component Cooling Water Heat Exchanger

Replacement

Bechtel calculation C-SJ-183-02,

CCW Heat Exchanger

Support Pedestal

Load

Evaluation,

Rev. 5, 1/18/89.

Quality Instruction JPN-QI-3.1,

Rev.

15, 7/89.

Bechtel Specification

5177-M900, Qualification of Seismic Category

1

Equipment, 4/87.

DEFICIENCY 89-203-13:

Re lacement

CCW Heat Exchan er Shell-Side

Nozzle Loads

Discussion:

FPSL engineering

package

PC/M 88-263,

was written for replacement

of the Unit 4 heat exchangers.

The heat exchangers

were qualified for the

imposed deadloads,

nozzle loads,

and seismic

loads in a Target Technology Ltd.

report.

Bechtel calculation M12-183-01 tabulated shell-side

nozzle

loads that were

substantially

higher than the nozzle loads, which Bechtel originally transmitted

to Target

and which were used in the Target qualification report.

Bechtel

transmitted

the revised nozzle

loads to Target

on November 23, 1988.

Bechtel

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...and Target, discussed

these

nozzle

loads

on December

1, 1988,

and Target

informed Bechtel that the increased

nozzle

loads

were acceptable.

However, Target never revised

and reissued

the

CCW heat exchanger

qualification

-, ...,, -.,report,,to

document the qualification of the heat exchangers

for the revised

nozzle loads.

This item is unresolved

(89-203-13).

Requirement:

10 CFR 50, Appendix B, Criterion III, states that design interfaces

shall

be

coordinated

and controlled.

References:

PC/N 88-263,

Component Cooling Water Heat Exchanger

Replacement.

Bechtel calculation M12-183-01,

CCW Heat Exchangers

Nozzle Loads Allowable,

Rev.

1, 1/12/88.

Target Technology report, Stress

Analysis of

CCW Heat Exchanger

Replacement/

Turkey Point Unit 4, Rev.

11, 10/4/88.

DEFICIENCY 89-203-14:

CCW Relief Valve

Re lacement

Discussion:

FPSL engineering

package

PC/N 86-238 was written for replacement

of Unit 4 relief valves

RV-1423 through 1431.

Teledyne

prepared

calculation

6548-1 to qualify these

1-1/2 inch or

1 inch diameter cantilever

branch lines

for both units.

Teledyne technical report TR-5322-1, requires that

safety-related

piping be qualified to the appropriate

loading combinations

and

stress

limits.

However, the Teledyne calculation

does not address:

The effects of valve thrust;

The need to support valve RV-3-1431 with a tieback from the 3-inch run

pipe, since the branch line is not rigid;

A stress

check at the root of each branch line for the combined effects of

pressure,

deadload,

valve thrust,

and seismic loads; or

The effects of lumped mass of the branch line and relief valve.

If this

lumped mass is considered,

the fundamental

frequency of the branch line

will drop.

This item is unresolved

(89-203-14).

Requirement:

10 CFR 50, Appendix B, Criterion,V states

that activities affecting quality

shall

be accomplished

in accordance

with the governing procedure

requirements.

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References:

PC/N 86-238, Safety Relief Valve Replacement.

Teledyne calculation 6548-1, Turkey Point Units 3/4 Component

Cooling Water

System Relief Valve, Rev. 0, 1/17/86.

Teledyne technical report TR-5322-1, Project Procedures

and Criteria/USNRC

I&E Bulletin 79-14 Analysis/Turkey Point Units 3 and 4, Rev. 1, 1/29/82

DEFICIENCY 89-203-15:

Com onent Coolin

Water 0 eratin

Procedures

Discussion:

The inspection

team reviewed operating

procedure

3/4-OP-030 for

the

CCW system to determine if valve lineups were consistent with approved

design

drawings

and if lineup changes for various operating evolutions were

properly established

and restored.

The team found the following deficiencies

when comparing the procedures

to system drawings 5610-T-E-4512,

Sheets

1 and 2:

Procedure

4-0P-030,

page 25, Step 7.5.2.7, directs the operator

to makeup

the

CCW surge tank as required

by manipulating valves t10V-4-832, 4-711B

and 4-710B.

Valve 4-737C was not referenced

as requiring manipulation;

however, it is shown as

a normally closed valve on the system drawing and

would inhibit makeup flow if not opened.

Procedures

3-OP-030

and 4-0P-030,

page 27, step 7.3.2.15 specify that

valve 4-711B be left open although drawing 5610-T-E-4512 indicates the

normal position as closed.

Step 7.6.2.15 specifies

the positions of

valves 3-711A and 4-711A as being left open although the referenced

drawing indicates

closed

as the normal position.

Procedures

3-OP-030

and 4-0P-030,

Attachments

2 and 3, for CCW valve

alignment inside and outside containment,

contained

the following

deficiencies:

Valves 3-10-681,

3-10-682,

and 3-10-683 were specified

as having

a

normal position of open in the valve lineup as

opposed

to closed

as

indicated

on the system drawing.

Valve 4-10-689

was specified

as being closed in the valve lineup as

opposed to closed

and capped

as indicated

on the system drawing.

Valves 3-10-749, 4-10-692,

4-10-1009,

4-10-1010,

4-1181, 4-1182,

and

4-769D were

shown

on the system drawings but were not included in the

valve lineup.

These deficiencies, if not corrected,

could have resulted

in failure to ade-

quately verify system integrity or could have placed the system in a condition

where valves were left opened

instead of closed, resulting in a loss of system

integrity.

With the exception of the makeup valves

(4-737C, 4-711B, 3-711A,

and 4-711A) the valves noted were vent and drain valves.

The licensee

agreed to make the identified corrections to the procedures.

This item is unresolved

(89-203-15).

B-5

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'ackages

identified the

need for additional

supports or modification of exist-

ing supports to bring stress

levels within the Final Safety Analysis Report

(FSAR) limits.

".="~'-,: ".-'""The'team was informed that

a functionality review of all packages

determined

that two supports

did not meet the stress

requirements for functionality and

that approximately

25 functionality problems

were identified that required

correciion of the piping or tubing systems.

The team determined that the

identified piping and tubing deficiencies

have

been corrected

and

no items

remain open.

The team was informed that

some supports

do not meet the

FSAR stress

require-

ments

and that an upgrade

program is scheduled

to be completed

by

December

1990.

Appendix

C identifies this as

an Inspector

Followup item.

2.8.3

Reactor Protection

S stem

The team reviewed the

DBD and verification report.

The team also reviewed

the Select Licensing Issues

(SLI) DBDs covering separation

and single failure

criteria, plant modifications,

system drawings, plant procedures,

and survei 1-

lance testing.

The team also performed

a walkdown of the control

room and

instrument rack room areas

to confirm proper

switch alignment.

In general,

the team was impressed

by the depth, breadth,

and quality of the

DBD and found it a useful

and well-researched

document for capturing the

design intent without being unduly encumbered

by configuration details.

However, the team concluded that the breadth of the

DBD verification was

insufficient in some cases.

The

DBD verification for the

was limited to

IEEE Standard

279 conformance,

which is a necessary

but incomplete criterion on

which to base

RPS functionality.

The licensee

did not include the verification

of critical

RPS functional requirements,

such

as response

times in the report.

The team was told that the licensee

took credit for periodic surveillance

test

results

and other plant programs to provide

a verification method for perfor-

mance attributes

such

as response

time and other design attributes,

such

as

separation.

The licensee

provided

a preliminary draft of a

DBD attribute

verification matrix for the

RPS.

This matrix identified references

to configu-

ration drawings,

performance

analyses

and calculations,

performance tests,

and

operability documents that verify key design attributes of the system.

It

appeared

that the verification matrix would enhance

the current

DBD

verification.

The team discussed

with licensee

engineering

and plant personnel,

the disposi-

tion of a variety of PWOs.

The

PWOs involved the

RPS fuse replacement

program

supported

by Westinghouse,

relay failures, the thermography monitoring program,

instrument vibration problems,

and various unrelated

items.

From these discus-

sions,

the team concluded that the

PWOs were properly dispositioned.

The team also identified one concern in the

CCW design involving a potential

single failure vulnerability that required

a successful

Failure trode and

Effects Analysis

(FMEA) to resolve.

Thq team noted that the

DBD validation

appears

to deliberately

exclude the use of Fh1EAs.

12

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2.8.4

Com onent Coolin

Water

S stem

The team reviewed the

CCW system design basis

documents,

plant procedures,

system performance calculations,

technical specifications,

system modifications

and vendor manuals.

The team also performed

system walkdowns.

Mechanical

S stems

To ensure that the required operator actions

were compatible with the plant

design,

CCW Operating

Procedures

(OPs) 3/4-OP-030

and Off-Normal Operating

Procedures

3/4-ONOP-030

and 3/4-0HOP-3108.2

were reviewed.

The team determined

that the valve lineups for procedure

3/4-OP-030

were incorrect.

The

CCW surge

tank makeup directions failed to include

a step to open valve 4-710B which

would be necessary

to achieve flow to the tank.

Several

vent and drain valve

positions

were specified in a position contrary to the system

TSE diagrams.

The identified concerns

could have resulted in a loss of system inventory.

This item is unresolved

(89-203-15).

The team also determined that procedure

3/4-ONOP-030 for loss of component

cooling water flow could not be implemented.

The Unit 4 charging

pump room was

not provided with emergency

hoses,

the hoses

in the Unit 3

pump room were not

of sufficient length,

and the hose couplings in the Unit 4

pump room were

corroded

and could not be used.

The licensee

provided dedicated

hoses of

sufficient length in each unit and replaced

the hose couplings.

Setpoint

change

79 which revised the standby

CCW pump start

on low header

pressure

from 75.0 psig to 60.0 psig was reviewed

by the team.

Logic diagrams

5610-T-L1, Sheets

24D and 24E,

showed

a setpoint of 78.5 psig.

The team was

informed that the diagrams

were in the revision cycle and that the control

room

prints

had been red-lined.

FP&L letter L-86-112 described

the activities associated

with the formation of

a

CCW valve task team.

The team was informed that

CCW valve operability was

examined

and that the identified issues

were documented

on the select

system

deficiency master punchlist.

Some of the valves required repacking or seal

replacement

to ensure

the abi lity to perform in the post-accident

alignment.

The most significant overhaul

was replacement

of the emergency

containment

cooler containment isolation valve liners to reduce

excessive

leakage.

In 1984, Westinghouse

notified the

NRC of a reportable

item associated

with a

potential overpressurization

condition in the

CCWS system

due to a closure of

the surge tank vent valve due to high radiation.

recommended

disabling the surge tank vent valve in the open position and removing the surge

tank relief valve.

The licensee

developed plant modifications for both units

in response

to the recommendations.

In 1987, the vendor subsequently

reported

to the

NRC and the utility that implementation of the earlier

recommended

modifications would result in a violation of containment isolation require-

ments.

The plant modifications were placed

on hold and the licensee

demonstrated

that the system could support the postulated

overpressurization.

In addition, the licensee

evaluated

the release

of radioactive

gases

through

the surge tank relief valve.

The team reviewed the licensee's

documentation,

procedures,

and decisions

and found no weaknesses

or concerns.

13

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Mechanical

Com onents

FSAR Appendix 5A designates

the

CCW pump and the

CCW surge tank as Class I

components.

As such, this equipment

was to have

been designed

to withstand

....,.,;...earthquake

and other specified

loads without loss of function.

The team found

that

FP&L could not access

the design documentation

to confirm that the

pump and the

CCW surge tank had been seismically qualified, or that an anchor-

age

check for either

component

had been performed.

FP&L was able to provide

the team with the Westinghouse

design specification

and

pump data sheet for the

CCW pump, which detail the

pump seismic design requirements.

The licensee

initiated

a review to regenerate

the equipment calculations.

This item is

unresolved

(89-203-16) .

The team reviewed the safety-related,

large bore piping analysis

performed

by

Teledyne for conformance

to FSAR requirements.

The Teledyne effort was

an

extension of previous work which Bechtel

performed to evaluate safety-related,

large bore piping for functionality as part of the

NRC Bulletin 79-14 review.

The team found that the Teledyne stress

packages

could not be audited since the

calculations

use information such

as valve weights

and offsets

and equipment

nozzle thermal displacements

without clear reference

to the originating Bechtel

calculations.

The Teledyne analysis

was not structured

in a manner to support

an independent

audit.

This item is an Inspection

Followup item (89-203-17).

The team reviewed

a sample of approximately

two dozen pipe support calculations

which Teledyne

prepared

to qualify the

CCW pump and

CCW heat exchanger

piping.

The following concerns

were identified:

The baseplate

procedure

used

by Teledyne

was not applicable to the stan-

chion configurations that were qualified for uplift loads;

Some anchor bolt tension

and shear

loads were not computed in accordance

with Teledyne's

baseplate

procedure;

Bending stresses

in single-angle

supplementary

support steel

were not

correctly computed;

The Teledyne stress

packages

indicated that seismic inertia and

ZPA loads

were

combined absolutely,

although these

values

were actually combined

by

the square root of the

sum of the squares

method for use in qualifying the

pipe supports;

and

The

beam stiffener American Institute of Steel Construction

web crippling

check was not performed.

These

concerns

are unresolved

(89-203-18).

The team performed

a limited review of the Bechtel

program to qualify

safety-related

small bore piping.

One Bechtel

walkdown package

and backup

calculation

was reviewed.

The governing Bechtel specification required that

the cantilever branch lines

be rigid and satisfy the applicable stress crite-

ria.

The calculation which the team reviewed accepted

two branch lines with

frequencies

that were not in the rigid range, without requiring additional

supports to stiffen the branch line or additional

analyses

to properly seismi-

cally qualify the branch lines.

This item is an Inspector

Follow-up item

(89-203-19).

14

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Desi

n Basis Documentation

The team conducted

a detailed review of the

CCW DBD, CDR, and the

DBD verifica-

tion report to assess

the adequacy of the

DBD reconstitution

program,

the

adequacy

and pertinence of the information contained

in the

DBD, the prioriti-

zation

and resolution of open

items and missing essential

documentation,

and

the degree of verification and validation of the information contained

in the

The team noted that Westinghouse

supplied the components,

designed

the

CCW system,

and

was tasked with preparing the

DBD documents.

The team observed that the

DBD contained

an adequate

level of system functional

basis

information and was well-supported

by reference

to-supporting

documents

and calculations.

The licensee

listed eleven calculations

performed during the

preparation of the

DBD which were associated

with the determination of the

component

and system heat

loads

and flow balance

requirements

to ensure

ade-

quate

performance of the system under normal, design,

and postulated

accident

conditions.

The team noted the following minor concerns with the system

DBD:

The list of components

that automatically isolate

on high containment

pressure

did not differentiate between

the Phase

A isolation signal

and

the Phase

B isolation signal.

The licensee will correct the list in the

next revision of the

e

The calculation references

were not consistent within the

The

licensee will correct this in the next revision of the

The

DBD stated that the system relief valves should be located at the

boundary of the

low pressure

section of the

CCW piping instead of on the

surge tank to accommodate

a break flow into the

CCW system during an

RCP

thermal barrier tube rupture event.

The licensee

noted that this was

a

recommendation

and not

a requirement,

and will clarify the statement

in

the next revision of the

The review of the portion of the

DBD that contained

the

CCW CDR led to many

questions

about the information contained in the document.

The team noted that

many of the stated

design

requirements

were not traceable

due to the lack of

references.

The document

was not well-prepared or reviewed.

It contained

information from the procurement specifications that did not reflect true

design

requirements

and information that was inconsistent with the system

design basis,

as illustrated by the following examples.

System configuration

changes

were designated

as pre-Operating

License

(OL)

and post-OL.

However, all of the configuration

changes

had occurred prior

to issuing the operating

license

on Unit 3.

The licensee

agreed to

correct the discrepancy

in the next revision to the

The

CDR implied that the original chemical mixing pot had been replaced

and that the design pressure

and design

temperature

of the

new pot were

indeterminate.

The licensee

reviewed the information and determined that

the

CDR statement

was erroneous

because

the original mixing pot was only

moved to a different location in the system.

The licensee

agreed to

correct the information in the next revision of the

15

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" 'he description of the

CCW pump start sequencer

was incomplete regarding

the

pump

A breaker configuration.

The licensee

agreed to correct

the

information in the next revision of the

The required stroke times for the subject

containment isolation valves

were identified as

10 seconds

at 125 psid.

The surveillance

procedure for

valve stroke times listed closure stroke times of 30 to 120 seconds

in

accordance

with the technical specifications.

The licensee

noted that the

CDR requirement

was

a procurement

requirement

and was not related to

accident mitigation requirements.

The licensee

indicated that it would

revise the

CDR to clarify the requirement.

The

RCP thermal barrier heat exchanger inlet check valves were identified

as subject to Inservice Inspection

( ISI) and testing requirements.

The

licensee

determined that the supplemental

testing

was not required.

The

licensee

agreed

to correct the information in the next revision of the

The design

parameters for the component cooling

pump suction pressure

instruments

were identified as

unknown.

The licensee

determined that the

statement

was erroneous

and agreed to delete it in the next revision of

the

CDR and to include the appropriate

information.

It appeared

that the licensee

had accorded substantially

less

importance to the

CDR document than the system

The information found in the

CDR was often

in error or inappropriate.

The licensee

subsequently

issued

a directive that

e

established

cautions

and limitations on the use of the

CDR information.

Therefore,

the

FPSL end-users

were encumbered

with verification of the

CDR

information before using it as design input.

The team reviewed the

DBD system verification that was performed

by

Westinghouse.

The report verified 34 design basis

statements

by verifying

performance

requirements,

basic functions,

and interface requirements

that

In some

cases

the verification was not completed

and

was reported for resolution.

The report also identified some areas of inconsistency

between the design basis

and plant documents.

The team concluded that the

CCW system design basis

requirements

were adequately verified and that the areas of incomplete verifi-

cation we'e appropriately identified.

The team noted,

however, that the

verification process

was primarily a documentation

review and very little

comparison of actual plant test results,

equipment

performance

and physical

walkdowns

was associated

with the verification process.

A notable exception

was the consideration of special test 86-05 involving the flow balance test

for the

CCW system.

The licensee

subsequently

generated

a

DBD verification

matrix that correlated

CCW testing with the performance verification

attributes.

A significant weakness

in the

DBD reconstitution

program was the lack of a

verification program for the information contained

in the

CDR document.

The

team was informed that

FPSL would perform an undefined verification of the

CDR

information.

This item is unresolved

(89-203-20).

An additional

weakness

in the verification process

was the exclusion of verifi-

cation for single failure vulnerabilities

as well as external

hazard

design

bases

such

as seismic, internally generated

missi le, compartment flooding,

16

I

~

fire, and electrical separation.

The licensee

informed the team that other

programs

(such

as Appendix

R

environmental qualification, high energy line

break,

and

IE Bulletin 79-14

sufficiently addressed

the design

bases for the

system

. The team determined that component seismic

response

and support

load

calculations

were not available for the

CCW pumps

and the

CCW surge tank as

documented earlier in this section.

A verification program for the hazards

would have identified this missing documentation.

The licensee

had prepared

a punchlist of open issues resulting from the verifi-

cation process.

Several of the punchlist items which related to emergency

containment

cooler heat removal performance,

excessive

CCW supply temperatures

under certain accident conditions,

and possible revisions to the

DBD heat

and

flow balance tables,

were addressed

in the SDI program described

in

Section 2.7.

Instrumentation

and Control

The team reviewed the

CCW design

aspects

related to leakage detection

and

mitigation.

The associated

instrumentation

was reviewed relative to the design

basis

leak rate,

the plant operating

procedures,

the surveillance

requirements,

and the operator

response

aspects.

The maintenance

instructions for the

surge tank level instrumentation calibration were found lacking in detail.

The

team was informed that

a procedure

upgrade

program for the instructions is in

progress.

The local

CCW surge tank level indicator for Unit 3 was found

inoperable

by the team

and

PWO 6571/63

was issued to fix the indicator.

The

team reviewed

FPKL Technical

Issue

76, which stated that

no credit is taken for

the surge tank level instrumentation

channel.

The analysis of a failure or

inoperability of the

CCW level indication channel

was also reviewed.

The

analysis

demonstrated

that the fai lure modes

and consequences

are within the

plant design basis.

The team identified several

errors

on the

CCW PODs

and

an instrument

loop

diagram:

I) The

CCW pump breaker

logic incorrectly

showed the breaker would

close

on

a loss of offsite power

(LOOP) load shed signal.

The correct logic

would block the

pump start during

a

(LOOP) load shed.

The elementary

diagrams

properly implemented the logic.

2) The diagram incorrectly depicted the

start/stop

selector

switches for the

CCW pumps

as maintained in the stop

position.

The switch was also found to be

a pull to lock in the stop position

with a spring return from the stop position.

The elementary

diagram correctly

depicted the switch characteristics.

3) The T-E diagram incorrectly showed

a

functional connection

between

the

CCW pumps

and the

pump discharge

pressure

indication channel

which was only a local gauge.

4) The termination contacts

for relay

PC-611X were inconsistently

shown

on the instrument

loop diagram.

The licensee initiated

a correction

and investigation of the drawing errors

and

this item remains

an Inspector Follow-up item (89-203-21).

The automatic

pump start circuitry for a low

CCW header

pressure

was reviewed

by the

team.

A common control relay was provided for all of the

CCW low

pressure

start circuits.

The relay contact output wiring was found bundled

together in the field.

The team reviewed

a

Ft1EA for the circuit configuration.

The

FHEA demonstrated

that internal rack wiring single failures would not

disable

both

CCW trains.

The

FMEA took credit for the ungrounded direct

current control circuit, such that

a single ground would not blow the fuses

and

that the ground faults are detected

and cleared.

The team noted that the

off-normal procedure for clearing grounds

does not require

an immediate action

17

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A

and does not specify

a time limit to allow the ground to exist.

This is an

u

~

inspector follow-up item to evaluate

the administrative controls to preclude

ground faults from existing for extended

time periods

(89-203-22).

The team concluded that the licensee's

design basis reconstitution effort was

adequate, in the areas

covered

by the program, but that overall weaknesses

existed in the verification and validation process

due to its limited scope.

In particular, the team also questioned

the licensee's

approach with respect

to

verification and validation of the

CDR,

and felt that effort should

be expended

by the licensee

to make it a reliable document.

The team also found

some minor errors in the instrumentation

and control

PODs,

but concluded that the errors would not have significant impact on safe

operation or design of the plant.

The team found several

deficiencies

and

weaknesses

related to a lack of correspondence

between

the operating

procedures

and the plant configuration,

and the improper staging

and poor condition of

emergency

equipment.

Supporting calculations to demonstrate

the

CCW pump and

surge tank load and anchorage

adequacy

were not available.

2.8.5

Electrical Distribution

S stem

The team reviewed the electrical distribution system including the

EDG loading

calculations,

the capability to start large motors,

load sequencing,

the

battery rating and testing,

the replacement

of station battery chargers,

and

the

EDG enhancement

project.

The associated

calculations

and drawings were

reviewed with respect to the technical

requirements

contained within the

DBD

and

FSAR.

The team found that the

DBD,

FSAR and drawings were in agreement.

However the

calculations

were not consistent with the

FSAR and

DBD, and in some

cases

were

not available.

The inconsistencies

included battery capacity,

the method of

calculating short circuit currents,

instrument voltage tolerances,

and

EDG

terminal voltage.

The team was informed that the

DBD and

FSAR would be revised

appropriately.

These

were further examples of CDR errors discussed

in

Section 2.8.4.

The short circuit analysis of the Class

lE buses

was found incomplete.

Short

circuit calculation 5177-203-E-02,

issued in 1983, indicated the need to

perform an additional calculation to include the contribution of the

EDG.

The

licensee

generated

the missing calculation

and the results

were satisfactory.

As a result of its review of the

EDG motor starting,

the team questioned

the

CCM pump load value and the extrapolation of the

EDG response

to the

480 volt

loads.

The licensee

performed

a

new analysis,

in response

to the team's

questions,

to demonstrate

that the restart of the

CCll motor could be accom-

plished

when the diesel is running in a loaded condition.

The motor voltage tolerance

was specified

as plus/minus

10 percent in the

DBD,

although plant operating

procedures

allow the

EDG voltage output to operate

in

the range of plus/minus

15 percent.

The team determined that the operating

procedures

were in error and that the

EDG voltage regulator actually controls

the voltage to plus/minus

one per cent.

18

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The team reviewed the adequacy

of the station batteries.

The minimum

end-of-service-life battery terminal voltage was stated to be

105 V.

The

licensee

has performed individual voltage drop calculations for load addition

or modification.

However,

no general tabulation of all loads exists to demon-

strate

the acceptability of the battery terminal voltage, i.e., that the loads

will function at 105 V.

The team reviewed selected test results that were

performed with the battery in a discharged

state.

Equipment appeared

to

function properly; however certain equipment

was bypassed

during the test.

Therefore,

the team remains

concerned that the testing

and available calcula-

tions'Mid not adequately

demonstrate

that the dc-powered

equipment

would

function at

105

V.

This item is unresolved

(89-203-23).

2.8.6

Plant Confi uration

The team performed

system walkdowns for the

CCW, instrument air,

and

systems.

Specific findings from system walkdowns

and general

observations

are

listed below:

e.

Instrument air control valve CV-4-2803 was erroneously

labeled

as

a

containment isolation valve.

The containment isolation valve label

was

removed to correct this problem.

During

a containment

walkdown for Unit 4, the team identified various

valves belonging to the breathing air, demineralized

water,

and service

air systems that were not labeled inside containment.

The team was

informed that the licensee

plans to label the service air and

demineralized

water systems

during the next outage.

A program has

been

implemented to tag the breathing air system during

a future outage.

The team identified a broken

gauge faceplate

inside containment for

pressure

gauge TI-4-1467.

The licensee

stated that it would replace

the

gauge.

The team identified two

CCW relief valves

on the containment

coolers that

had tape covering the vent ports.

The licensee

subsequently

removed the

tape.

Excessive boric acid crystal buildup was identified around the valve

bonnet flange on two containment

spray

pump suction valves.

The licensee

informed the team that the valves would be fixed during the next outage.

As discussed

in Section 2.3,

some discrepancies

were found between

the plant

drawings

and the as-built plant configuration.

The assessment

of these errors

was that plant operability would not be affected.

During

a general

area

walkdown the team observed that 4-inch holes

had been

drilled into the bottom of the lagging covering the electrical

cables to the

safe

shutdown

panels at 3-inch intervals.

The licensee

stated that water and

high humidity problems

caused

the lagging to swell and

bow, and that drainage

holes were drilled in the bottom of the lagging to allow water to escape.

The

team reviewed Industrial Testing Laboratories report 86-1-143,

which demon-

strated that an equivalent

lagging configuration

had passed

the ASTN-E119 test

requirements

to limit the inner conduit surface

temperature

to 325 degrees

~

Fahrenheit.

19

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hith the exception of the minor concerns

noted above,

the plant configuration

~

was found satisfactory.

2.9

Surveillance Testin

The team reviewed the

CCW surveillance

procedures

and applicable

TS require-

ments.

The team identified that the monthly flowpath verification procedure,

3/4-0SP-030.3,

did not address

the

CCW pump seal valves

and the

A and

8 header

cross-connect

valves.

The licensee

incorporated

both sets of valves into the

lineup verification procedure.

The

TS requires that the emergency

containment

coolers

be tested

during each

refueling outage to ensure

2000

gpm flow through

each cooler unit.

The appli-

cable surveillance

procedure,

3/4-0P-4704.5,

described flow test performance

with two pumps in operation.

However, the post-accident

lineup is expected

to

include only one

pump.

The team was informed that

a

CCW system flow balance

had been

performed in 1986

and that the system valves were aligned for proper

single-pump operation.

The licensee

stated that

a correlation

between

the one-

and two-pump operation

would be incorporated

into the revised

TS to remove the

current ambiguity.

The team reviewed the IST program for containment isolation valves.

Review of

the

DBD CDR indicated that it specified the

CCW containment isolation valve

stroke time as

10 seconds.

The

IST stroke times for the valves were

30 seconds.

The team was informed that the

DBD CDR time limits were based

upon

vendor recommendations

and that the

DBD CDR would be revised for consistency

with the IST requirements

which implement the applicable

TS requirements.

The team noted that FP8L's Measuring

and Test Equipment

(NOTTE) program was very

comprehensive

and technically acceptable.

The team questioned

the controls for

loaning ATE to plant employees for personal

use

and

was informed that the

equipment

would be calibrated

both prior to leaving the site and upon return.

The team found the surveillance test procedures,

particularly the Integrated

Safeguards

Test

( IST), technically adequate.

However, maintenance

calibration

procedures

did not contain sufficient details for the

CCW surge tank level

instrument calibration.

FPSL informed the team that upgraded

maintenance

calibration procedures

were being developed similar to the in-place surveil-

lance program.

The team also reviewed the

EDG surveillance testing

and found that the test

results

under simulated

loadings were not extrapolated

to predict the

EDG

response

to the actual

loss-of-coolant-accident

(LOCA) loading.

The team

reviewed the

EDG vendor, startup,

periodic loading,

and sequencing

test

results,

and concluded that the

EDG has been'roperly

tested.

The response

of

the

EDG was questioned

during the situation

when the

EDG is paralleled to the

off-site source during the test.

The team reviewed the

EDG response,

which

would include isolation from the off-site source

and mitigation of the loss of

offsite power.

The successful

response

would be achieved with an assumed

single failure and without violation of redundancy

between

the two redundant

trains.

The team reviewed the fuel oil transfer testing

and noted that the procedure

valve nomenclature

did not correspond with the plant valve identification.

Due

to

a relabeling of equipment,

discussed

in Section 2.2.1,

the test procedure

did not reflect the current

component identification.

The team confirmed that

20

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the proper equipment

was operated

during the test.

The licensee

implemented

a

procedure

change to correct the improper valve nomenclature.

The team reviewed the ammeter

instrument accuracy with respect to the battery

surveillance test.

To account for the instrument accuracy,

FP8L documented

that sufficient margin existed

between

the actual test loading and the duty

cycle presumed

in the battery calculation.

The team also questioned

the

procedure

sequence for checking

and tightening the battery cell cable

connec-

tors.

FP&L revised the test procedure to preclude

these

concerns for future

testirfg.

The calibration of the source-range

nuclear instrumentation

was reviewed

by the

team.

The procedure relies

upon megger tests.

However,

megger testing

measures

only impedance

to ground

and cannot

be relied upon to detect

cable

degradation.

Other attributes

such

as conductor

impedance

need to be validated

before

a comprehensive

conclusion

can

be reached

regarding

cable degradation.

The use of time-domain reflectometry

(TDR) is an alternate

system that

can

identify problems with the cable

and connector

system.

The

TDR system

compares

reflected versus original pulses to identify weak links in the system.

Based

upon the team's

review, the licensee

has properly structured

the survei 1-

-.lance program to verify the operability of the system

components.

2.10

Maintenance

0

The team reviewed the predictive maintenance

program that utilizes vibration

analysis, oil sampling

and thermography

techniques,

as well as the following

documentation

regarding

CCW system maintenance activities:

corrective

and

preventive maintenance

procedures;

Plant Work Orders

(PWOs); Nuclear Plant

Reliability Data System

(NPRDS) failure information; and vendor technical

manuals.

The team's

review of

CCW maintenance

procedures

and the associated

equipment

vendor manuals

found them satisfactory with regard to format, incorporation of

vendor requirements,

quantitative

acceptance

criteria, coordination with

operations,

and post-maintenance

testing.

The team reviewed the licensee's

analysis-based

preventive maintenance

program

which included the predictive maintenance

program.

Routine monitoring of plant

equipment

has

been

performed which utilized vibration analysis, oil sampling

and infrared thermography.

The results

were trended

and analyzed to obtain

early detection of equipment degradation.

The

CCW pumps were examined

on a

quarterly basis.

The program has successfully

identified problems in other

plant systems,

such

as improper oil in a auxiliary feedwater

pump governor

and

electrical

system hotspots.

The team reviewed the

CCW component failure information.

The repetitive

pump mechanical

seal failures were found rectified by revising

a maintenance

procedure that had specified the incorrect seal

spring compression

value which

could have resulted

in increased

seal

wear.

The team reviewed the outstanding

PWOs for the

CCW system.

While the work

prioritization appears

proper,

157

PWOs remain outstanding.

Three of the

PWOs

have

been outstanding

since

1986, while the majority (109) were issued in 1989.

21

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RPS reactor trip switchgear, for example,

had several drafting errors.

~

Mith those exceptions,

the team found the

PODs in generally satisfactory

condstion.

Other drawings that exhibited

some problems are addressed

by

various drawing update

programs

which the facility is pursuing aggressively.

2.4

Desi

n Inte ration Review Team

A Design Integration

Review Team (DIRT) was established

by

PEP Task

6 to

coordinate

the various engineering disciplines during the design

and

design review processes.

The team reviewed the DIRT charter,

which stated

additionally that the DIRT's mission

was to review PC/Ms, safety evaluations,

and justifications for continued operations

(JCOs).

The team reviewed meeting minutes from a recent

DIRT session.

In addition, the

team attended

a DIRT meeting

and found good communications

among the corporate

and site engineering

organizations.

The DIRT discussed

a recently

issued

design equivalent engineering

package

(DEEP) and

a St. Lucie plant trip.

The

team found the DIRT process

functioning satisfactorily.

2.5

En ineerin

Desi

n Packa

es

As part of PEP Task 7,

FPSL issued

several

upgraded

procedures

to implement

enhanced

control of engineering

packages

(EP).

The inspection

team reviewed

these

procedures

and determined that they defined the responsibilities of

design organizations,

control of design interfaces,

release of design output

documents,

changes

to approved output documents,

changes

to drawings,

and

control of design records.

An EP is developed

in stages

in coordination with plant operations

and instal-

lation personnel.

The final design

package

may be audited

by the DIRT which is

described

in Section 2.4.

The licensee

uses

an

EP feedback sheet to provide

constructive

feedback to the design organization

regarding the quality of the

design output.

The team also reviewed the procedure for DEEPs, which are

issued to implement

a change that does not alter the plant design basis.

Typically, a

DEEP is used for like replacements.

2.5.1

E1 ectri ca 1

To evaluate

the facility's electrical engineering

design

packages,

the team

reviewed

PC/M 88-350 for the installation of the

new battery chargers.

Although the package

had not been fully approved, it had been design verified.

The purchase

specification for the

new chargers

had been

issued for bids,

and

two chargers

had been

shipped to the site.

The team also reviewed calculations

EC-051

and

EC-52 for battery charger sizing

and supply cables,

respectively.

The battery charger sizing calculation

was

found adequate.

The supply cable calculation,

however,

was found deficient

since the licensee

had not checked

the short circuit capability of the

cables.

The supply cables

were rechecked

by the licensee

and verified to be

satisfactory for service.

The licensee .stated that it would institute

a

formal procedure to preclude

a recurrence

of this problem in the future.

The team raised

the following concerns relative to battery charger

purchase

specification

FLO 53-20.2006.

II

'1

V I

c.

0

The specification failed to reference

IEEE 650-1979 for qualification of

the battery chargers

and did not require compliance with this standard.

The licensee

stated that, after bid issue

and at the suggestion

of the

vendor,

compliance with IEEE 650 was

added

as

an amendment to the original

specification.

This should

have

been specified

by FP&L, since the design

intent was to comply with current industry standards.

T)e specification stipulated

an adjustable

charger equalizing voltage

range

between

2.33

and 2.38 volts per cell, which would result in a

battery voltage of 139.8 to 142.8 volts.

Plant operating

procedure

O-SME-003.3 defined the maximum equalizing voltage

as

138 volts.

The

licensee

confirmed that the charger

specification

was incorrect

and would

be revised.

In summary, while the team found the

PC/M adequate,

the associated

specifica-

tion and calculations

lacked proper coordination,

and did not contain clear

definition of the applicable requirements,

and contained technical errors.

2.5.2

Instrumentation

and Control

The team reviewed the following modification packages:

PC/M 86-239,

RPS Test Selector Switch Replacement

PC/M 87-114,

CCW Heat Exchanger

Thermometer

Replacement

with RTDS.

PC/Ms86-239

and 87-114

concerned

replacement

of safety-related

components

~

within the

and

CCW systems

respectively.

Both provided thorough

and

explicit safety evaluations,

clearly stated

the bases for the modification,

and

provided explicit design requirements.

PC/M 86-239 indicated that the facility properly consiaered

the effects of any

new failure modes or effects introduced

by the

new configuration.

The team

examined portions of the revised surveillance test with respect

to the revised

wiring diagrams,

annunciator

drawings,

and other design

documents

and found the

implementing documentation

to be consistent.

PC/M 87-114 explicitly documented

requirements for channel

accuracy

and

provided sufficient engineered

design details for proper procurement

and

installation.

The team asked

the licensee to verify that the associated

flow

channels

were designed

in a manner that would promote sufficiently accurate

and

repeatable

measurements.

The licensee

demonstrated

that adequate

upstream

and

downstream conditions

had been provided for the flow elements.

During the review of drawings associated

with PC/M 86-239, the team identified

several

errors

on the

RPS wiring and connection

diagrams.

These

involved

incorrect status light wiring, improperly identified switches,

poor cross

references,

and ambiguous

terminal identification.

The licensee

reviewed the

drawings

used for work implementation

and determined that the identified

concerns

stemmed

from drafting errors.

The licensee

agreed to correct the

affected drawings.

The team noted that the requirement for failure mode evaluation

and explicit

safety evaluations

was reflected in the modifications.

The team also noted

that the concept of the engineering

package

feedback

sheet

provides

a good

I I

tl

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'it

~

mechanism to solicit feedback

on the quality of the engineering

package.

The

team was impressed

by these

two recent modifications.

The packages

provided

thorough

and explicit safety evaluations,

clearly stated

the bases for the

modifications,

and provided explicit design requirements.

2.5.3

Mechanical

Com onents

The team reviewed the following design modification packages

concerning

mechan-

ical equipment modifications:

PC/M 88-263,

Component

Cooling Water Heat Exchanger

Replacement;

and

PC/M 86-238, Turkey Point Unit 4 Safety Relief Valve Replacement/Component

Cooling Water System.

A Bechtel calculation,

prepared to check the concrete

pedestals

supporting the

heat exchangers,

erroneously

concluded that the heat exchanger

was.,rigid with a

fundamental

frequency greater

than

33 Hertz (Hz).

Subsequently,

Target Tech-

nology performed

a frequency

check and confirmed that the heat exchanger

shell

is rigid.

Target then used the safe

shutdown earthquake

(SSE) zero period

acceleration

(ZPA) to qualify the heat exchanger

shell

and attachments

based

on

the fundamental

frequency calculated

by Bechtel.

The team found that the

Bechtel calculation failed to consider the transverse flexibilityof the

concrete

pedestals,

which would affect the heat exchanger

response

and would

lower the fundamental

frequency to approximately

10 Hz.

As

a result,

the

ZPA

loads

computed

by Target Technology would represent

less

than half of the

seismic

loaas that the heat exchanger

would engage

based

on the flexibilityof

the concrete

pedestals.

Consequently,

the Target stress

report does not

adequately qualify the replacement

CCW heat exchangers

for the governing

seismic

loads.

The licensee initiated

a review of the calculation.

This item

remains

unresolved

(89-203-10).

Teledyne

prepared

the stress

analyses

to qualify the shell-side piping of the

CCW heat exchangers.

Teledyne

modeled the circumferential

and longitudinal

nozzle stiffness

as rotational springs,

and restrained

the remaining rotational

degree of freedom and the three translational

degrees

of freedom.

Because of

the concrete

pedestal flexibility, the translational

degree of freedom at each

shell-side

nozzle-piping interface

should also

be released

and modeled

as

a

translational

spring.

Adding this translational

spring constant to the mathe-

matical piping model

may affect the frequency

response

and the stress

magnitude

of the attached

piping.

The licensee initiated

a review of the calculation

and

this item is an Inspector Follow-up item (89-203-11).

The team reviewed the Bechtel calculation which was prepared

to check the

heat exchanger

concrete

pedestals

due to the heat exchanger reactions.

The

Bechtel calculation

makes

assumptions

regarding

the pedestal

steel reinforce-

ment because

the installation drawings

could not be found.

Consequently,

the

team could not check the following attributes:

the load transfer

between

the

heat exchanger

and the top of the pedestal;

the load transfer through the

pedestal

anchor bolts; the structural

capacity of the pedestal itself; and the

load transfer

between

the base of the pedestal

and the building concrete slab.

This item is an Inspector Follow-up item (89-203-12).

The team reviewed another

Bechtel calculation which tabulated

CCW heat

exchanger

shell-side

nozzle

loads substantially

higher than the shell-side

~ e

I

0

4'P

V

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t

~rf

P

nozzle

loads that Bechtel

had originally transmitted to Target Technology.

Although Bechtel transmitted

these revised

loads to Target Technology, Target

Technology never revised

and reissued

the seismic qualification report for the

CCW heat exchangers

to document the acceptability of the increased

shell-side

nozzle loads.

The licensee initiated

a review of the calculations

and this

item remains unresolved

(89-203-13).

The team reviewed the calculation prepared

by Teledyne to support

PC/H 86-238

for the replacement

of the Unit 4

CCW relief valves.

The calculation did not

address

the effects of valve thrust,

the need to support

a flexible branch

line, and the performance of a stress

check at the root of each branch line.

The licensee initiated a review of the calculation

and this item remains

unresolved

(89-203-14).

The team concluded that the documentation

associated

with recent Unit 4