ML18005A369

Section

3.8.3.2.

Violation 400/87-41-02,

Uncontrolled Change to Design Input

q P

The licensee

did not identify as proprietary

any of the material

provided

to or reviewed

by the inspectors

during this inspection.

3.

Licensee Action on Previous

Enforcement Matters

(Closed)

Violation item 400/86-69-01,

Failure to Protect

Permanent

Plant

Equipment.

This violation involved

craftsmen

using

safety

trays

as

scaffolding

and

attaching

scaffolding

to

permanent

safety-related

equipment.

The

licensee's

corrective

actions

for this violation are

stated

in their

December ll, 1986

response

to

NRC Region II for Inspection

Report

No. 50-400/86-69.

The cause of this violation was that the two craft personnel

who were

using

cable

tray

L1901

SR4

as

scaffolding failed to follow Work

Procedure

WP-48.

To correct this problem,

personnel

and construction

materials

were

removed

from the cable tray and the tray and installed

cables

were

inspected

for damage.

No

damage

was

noted.

The

two

craft

personnel

and their

foreman

were

reprimanded

and

the

construction

manager

issued

a

memorandum

dated

September

8,

1986 to

the project craft supervisor reiterating the requirements

of WP-48.

Regarding

the scaffolding which was attached

(wired) to an instrument

tubing track

and

was in contact with an

HVAC valve actuator,

the

licensee's

corrective

actions

included relocation of the scaffolding

and inspection of the track and actuator for damage.

No

damage

was

noted.

The

foreman

responsible

for erection of the scaffolding was

reprimanded.

b.

The

inspectors

examined

CP&L procedure

number

CMP-4,

Rigging

and

Temporary

Loads,

and

procedure

number

MMM-022, Rigging

Loads

from

Permanent

Plant Components.

These

procedures

were written to replace

procedure

WP-48 which was

a construction

procedure

and control

the

attachment

of rigging

and

temporary

loads

to

permanent

plant

equipment.

Violation 400/86-69-01 is closed.

(Closed)

Unresolved

Item

400/86-69-03,

Evaluation

of Containment

Bui'1'ding

Load

Reversal

Design

Error

as

Nonconformance

and

for

Reportability

to

NRC.

During

the

special

inspection

conducted

S'eptember

4-11,

1986,

documented

in

Inspection

Report

Number

50"400/86-69,

licensee

design

engineers

informed the inspector

that

errors

were

discovered

in

the

original

EBASCO

design

for

the

containment

building

(CB)

platform

steel.

Unresolved

Item

400/86-69-03

was identified by the inspectors

as

a result

of the

fai lure of the licensee

to write an

NCR to document

and disposition

this problem.

Subsequent

to inspection

86-69,

the

inspector

held

further discussions

with licensee

engineers pertaining to the final

design

verification

program

for the

CB platform steel.

These

discussions

disclosed that the final design verification program

was

a

systematic

in-depth

review of the

EBASCO design

and

was

being

conducted

to

include all

final

attachment

loads

(e.g.

piping,

equipment,

conduit cable trays, etc.) which were not in the original

analysis.

The

purpose

of this

review

was

to verify that

the

"as-built"

CB platform steel

structures

conformed to

and

NRC

design

requirements,

and

to modify the

CB steel

structures

when

necessary

to

restore

the

original

design

margin.

The

discussions

also

disclosed

that

the

licensee

only

suspected

the

problem concerning

the

load reversals

during inspection

86-69.

The

extent of the

problem

was

not identified until late September

1986,

after conferring with the

EBASCO design

engineers,

and

performing

some

additional

design

verification studies.

At that

time,

86-600

was written to disposition

the

problem.

The

CB steel

design

verification program

and disposition of

NCR 86-600

was tied to

an

oral

commitment to re-inspect

on the connections

for the

92

heaviest

loaded

beams

on the

CB platform.

This commitment

was

made

by

a

licensee

representative

during

the

September

4-11,

1986

inspection.

These

beams

were

defined

as

those

loaded

to 0.85 or

greater of the

code

allowable

stresses

(Note:

In Inspection

Report

Number 50-400/86-69,

this definition was incorrectly written that the

heavily

loaded

beams

were those

loaded at 0.75 or greater of the code

allowable

stresses.

The figure of 0.75 is in error.

The correct

figure should

have

been 0.85.)

During the licensee's

in-depth review

of

the

CB platform

steel

design,

licensee

engineers

identified

several

questions

regarding

design methodology.

These questions

were

documented

and

resolved

in

memoranda

and

various

design

calculation

books.

The

inspector

reviewed

portions

of the

CB

structural

steel

design calculations.

The calculations

reviewed are

discussed

in paragraphs

5.b and S.c below.

Based

on the discussions

with licensee

engineers,

review of various

design

calculations,

and

other

documentation,

and

review of the

sequence

of events,

the s"nspectors

concluded that

NCR 86-600

was not

reportable

under

10 CFR 50.55(e),

and

that

the

licensee

acted

properly in delaying initiation of an

NCR unti 1 late

September

when

NCR 86-600

was issued.

Unresolved

Item 400/86-69-03 is closed.

(Open)

Violation Item 400/86-77-01,

Failure

to

Implement

Adequate

Design Control Measures.

The licensee's

corrective actions for this

violation

are

stated

in their July 2,

1987,

August 4,

1987,

and

October

30,

1987

responses

to'RC

for

Inspection

Report

No.; 50-400/86-77.

The

August 4,

1987

response

corrected

some

statements

in

the

July 2,

1987

response.

The

October

30,

1987

response

extended

the date for completion of corrective action from-

November

1987 to March 31,

1988.

The cause of the violation was

due

to errors

in

design

assumptions

made

for distribution of weld

stresses

under specific connection

types

and loading conditions.

In

addition,

a. supervisor

was

involved

in

the

design

verification

process

for the calculation

for

FCR-AS-10360

in violation of the

licensee's

design

control

procedures.

The licensee's

actions

to

correct this violation involved review of the specific calculations,

i.e.,

Detail

G,

and

FCR

AS-10360,

which contained

the errors

in

design

assumptions

identified in the violation.

This

review

was

accomplished

during the inspection

and resulted

in issuance

of

NRC

Nos.

OP-86-0183

and

OP-86-0185

to

document

and

disposition

these

design

errors.

The specific errors

were corrected

through issuance

of field modifications for Detail

G and

a plant

change

request

for

the

spent

fuel

rack

support

design

(FCR

AS-10360).

Further

corrective

action

included

review of civil-structural calculations

for the

CB platform steel,

the

RAB

248 platform steel

and

steam

generator

lower lateral

supports.

This

review

has

been

completed

within the

corporate

Nuclear'ngineer

Department,

Civil-Structural

Design

Section.

However,

due

to

other

concerns

identified

by

individuals within the

CP&L organization

regarding

these

same design

calculations,

the

licensee

has

decided

to

conduct

an

independent

review of these

calculations

and

others

which involve structural

steel

design at the Harris Plant.

NCR Nos.

OP-86-0183

and

0185

are

still

open

pending completion of this independent

review.

The scope

of the

independent

review is outlined

in

a

CP&L memorandum

from

R. A. Watson

to

A. B. Cutter

and

HE

R.

Banks

dated

September

15,

1987, Subject:

Independent

Review of Structural

Steel

Design Issues.

This

memorandum

summarized

the areas

of concern

expressed

by various

individuals within the

CP&L organization.

The

concerns

are

as

fol 1 ows:

(1)

Potential

errors

in

RAB

248

platform

steel

calculations

identified in CP&L Letter No. MS-876288(E),

dated June

22,

1987.

(2)

Effect of addition of attachments

to the

steam generator

lower

lateral

supports.

(3)

Resolution of previously identified design deficiencies

in the

CB

platform

steel

calculations.

Also,

the

resolution

of

questions

raised

by HPES engineers

during review of the platform

steel

design.

The

questions

regarded

the

original

EBASCO

'design.

(4)- Possible

discrepancies

in the

HPES civil program for cataloging

,j:;-";-",'.and analyzing the effects of additional

attachments

to various

> -"..'-;:l':steel

structures.

(5)"'ossible

improper

FCR/FM justifications performed

by the

HPES

civil unit,

The inspectors

discussed

the

independent

design

review program with

the Manager of the Nuclear Engineering

Department.

These discussions

disclosed

that

an outside

consultant will be

retained

by

CP&L to

perform

an overall

review of the structural

steel

design

to assess

proper methodology in relation to AISC Code compliance.

In addition,

a

detailed

review will

be

conducted

by

two

senior

structural

engineers

who were not involved in the Harris structural

steel

design

work.

These

individuals are

independent

and

do not work under the

supervision of the principal engineer

who was formerly the Hanager of

the

HPES civil unit.

The results of the Independent

Review of the Harris Structural

Steel

design

issues will be examined

by

NRC Region II.

Pending

completion

of this review,

and closeout

of

NCR OP-86-0183

and

0185, Violation

Item 400/86-77-01 will remain

open.

(Open) Violation Item 50-400/86-77-02

Undersized

Welds

on Cable Tray

Supports.

The licensee's

corrective

actions for this violation are

stated

in their July 1,

1987

response

to

NRC for Inspection

Report

50-400/86-77.

The

cause

of the violation was due to failure of the

welders to provide acceptable

weld profiles

on the identified Detail

G welds

and fai lure of the

QC inspector s to identify the undersized

The licensee

issued

Nonconformance

Report

numbers

OP-86-0149

and

OP-86-163

to

document

and di sposition

the

undersized

Detai 1

G

During construction

of the Harris Plant,

the licensee's

QA

staff issued

several

NCRs regarding

undersized

These

NCR were

examined

by a Region II QA specialist during the September

4-11,

1986

inspection.

(Note:

For

a listing of these

NCRs

see

Inspection

Report

Number

50-400/86-69.)

An indepth

review of

the

licensee

corrective

actions

pertaining

to identification

and correction

of

undersized

structural

steel

welds will be conducted

by

a

Region II

welding

specialist

(i.e.

Metallurgical

Engineer

or

an

NDE

specialist).

Pending

completion

of this

review,

Violation

Item

86-77-02 will remain

open.

4.

Unresolved

Items

Unresolved

items were not identified during this inspection.

5.

Case

RII"87-A"0086

Background

An,individual, hereinafter referred to as the alleger,

contacted

NRC

Region II and expressed

several

concerns relating to undersized

on',;,structural

steel,

structural

steel

design

methodology

and design

of;;-; welded

pipe

. attachments.

Followup

on

these

allegations

is

discussed

in paragraphs

5.b through 5.d below.

Deficiencies in Structural

Steel

Design

(1)

Concern

Design deficiencies

were previously identified by

NRC on major

structural

calculations.

Other

structural

steel

design

deficiencies

were also identified by licensee

engineers 'in

an

internal

memorandum

dated

June

22,

1987.

In addition,

there

are other questions

regarding

design

of the containment

building structural

steel.

These

concerns

are listed

in the

table below.

TABLE

CB Platform Steel

Design Concerns/Contentions

Concern

Number

Contention

'onnections

were

modeled

in

the

computer

analysis

as pinned connections

when in fact

some

were

as

built

as

rigid

(fixed)

connections.

Stress

calculations

were not performed for

angles

and plates in some connections.

Load reversals

not

summed correctly.

Eccentricity not considered

in connection

or

weld design.

Wrong load case

chosen

as most critical.

Torsional

loads not considered.

Whip restraint'loads

not considered.

Expansion

joints

used

to limit thermal

stresses

are not functional.

The value for accident

thermal

temperature,

i.e.

hT = 148'F,

was not justified.

10

Use

of

1. 1

interaction

factors

were

not

justified.

Reduced

horizontal

"g" values

were applied

to vertical

loads in seismic analysis.

12

Supplemental

steel

was

added to

CB platform

to stiffen

member

and

reduce

weak

areas

stresses.

However,

the

loads

from

the

supplemental

steel

were

not

included

in

analysi s.

13

Hanger foot print loads were not resolved to

beam centroids,

14

15

Effect of

DBE loads

from pipe

supports

on

platform steel

not checked.

l

Use

of rigid plate

theory

used

to

check

embeds is not justified.

Discussion

The

design

deficiencies

discussed

by the alleger

which were

identified

by

NRC

are

those

associated

with Violation

Item

400/86-77-01.

These

deficiencies

involved

the

Detail

G

connection

on

Drawing

number

CAR 2168-G-251-501

(containment

building cable

tray

frames)

and incorrect application of the

AISC Ultimate Strength

Method for welds in design of connections

for the

new fuel pool racks covered

by

FCR AS-10360.

As stated

in

paragraph

3.c

above,

the

licensee

issued

numbers

OP-86-0183

and

OP-86-0185

to

document

and disposition

these

errors.

The

RAB

248 platform steel

concerns

are

documented

in

Letter

No. MS-876288(E),

dated

June

22,

1987.

This letter is

discussed

in Paragraph

3.c above.

The

15 specific concern/contentions

regarding

the

CB platform

steel

were identified by

CP&L structural

engineers

employed in

the Harris Project

Engineering Staff Civil Structural

Unit in

Fall

1986,

during

the

final

design

verification of

the

CB

platform steel

design.

These specific questions

were

addressed

to

the

Harris project Architect-Engineer,

EBASCO in

a

hand

written

telecopy

transmitted

to

EBASCO

at

4:03

p.m.

on

October 20,

1986.

The

inspectors

reviewed

a

copy

of thi s

telecopy

and noted that the alleger's

concerns

are

verbatim

as

those listed in the telecopy.

The

EBASCO response

to the

questions

is

documented

in

an

EBASCO

'memo

to

Lee Williams/Ron Knott

(CP&L

HPES)

from

T. McCarthy

(EBASCO)

dated

January

6,

1987,

Subject:

Harris

RCB

Platform

Supplementary

Analysis.

Additional

responses

to address

these

concerns

were

prepared

by the

CP&L Design

Engineers

to cross

reference

responses

to

the

concern/contentions

to

specific

EBASCO or

CP&L design

calculations.

This cross

reference

is

outlined

as

an undated,

unlabeled listing of 16 contentions

and

specific

response

(including

referenced

calculation)

which

addressed

the contentions.

The cross

reference

is included in

the

CB structural

steel calculation books.

In response

to this concern,

a review of the calculations

for

the

containment

platforms

was

conducted

by

a

Region II

structural

engineer

to

make

a

more

detailed

evaluation

concerning

the

adequacy

of

the

CB platform

steel

design.

Specif'ically,

several

highly-stressed

beams

(as

determined

by

CP&L and

EBASCO to have Interaction

Equation

values

exceeding

0.85)

were

checked

to verify design

procedures,

including

assumptions,

load combinations

and factors,

allowable stresses,

and structural

configuration.

Also, the calculations

for the

connections,

including welds, clip angles,

and

embedded

plates,

of

the

respective

beams

were

examined.

The

beams

checked

included

514,

515,

517 of the

Load Verification Calculation,

LV54,

Volume 4.

In

addition,

calculation

number

CAS15,

"Containment

Building

Platform

Static

Analysis

- Revised

Stresses,"

which is

a compilation

of

beams

which

EBASCO

was

unable

to

qualify

when

using

the

original

design

load

assumptions,

was reviewed.

(3)

The

preliminary

conclusion

reached

upon

completion

of this

evaluation

is that

the calculations

have

been

performed

in

a

manner consistent

with professional

standards

of practice

and

satisfy

design

and

requirements.

However,

additional

review of the structural

steel

design

concerns will be conducted

by

NRC

Region II in closeout

of Violation Item 400/86-77-01,

following completion

of the

licensee's

independent

review of

structural

steel

design

concerns

discussed

in paragraph

3.b.

Findings

The

concerns

expressed

by

the

alleger

had

been

previously

documented

by the licensee

in various internal

CP&L documents,

or had

been

previously identified by

NRC.

These

concerns will

be

further

evaluated

by

NRC

following

completion

of

the

independent

review being conducted

by the

licensee,'uring

review of calculation

LV-54, the inspectors

noted that

thermal

loads were'valuated

using

methods

specified

in

ANSI

N690-1984,

Steel

Structure - Design.

This design

is based

on

steel ductility and local effects of thermal

stresses.

During

review of the

FSAR, the inspectors

noted that this ANSI standard

was

not referenced.

The

inspectors

discussed

with licensee,

engineers

the

need

for

referencing

this

standard

in

the

appropriate

FSAR Section

(Section 3.8.3.2)

when

the next

'.'j.:; amendment is submitted to

NRC the Licensee.

This was identified

-",to the licensee

as

Inspector

Followup Item 87-41-01,

Add ANSI

';:.,;,...;N690-1984

ta

List

of

Applicable

Codes,

Standards

and

Specification in

FSAR Section 3.8.3.2.

c.

Undersized

Melds on Structural

Steel

Connections

Concern

Compounding

the

concerns

discussed

in

Paragraph

5.b.,

above,

there

are

an

unknown

number of undersized

on structural

steel

connections.

As

an

example,

the

alleger

stated

that

undersized

were

found in

126 of 187 connections

examined

on the

CB platform steel.

The connections

had to

be modified

because

of the undersized

(2)

Discussion

The

licensee

examined

92 of the

highest

stressed

beams

for

possible effect of undersized

This was in response

to

concerns

expressed

by

NRC Region II during the

September

4-11,

1986

inspection

which

are

documented

in

Inspection

Report

Number 50-400/86-69.

In addition to these

184 connections

(92

beams

x

2 connections

per

beam

= 184),

the

licensee

evaluated

three

other individual connections

which were considered

to be

critical and highly stressed

(Note:

Highly Stressed

is defined

as

a

beam

loaded to greater

than 0.85 fy).

In order to perform

the evaluation,

a detailed as-built sketch

was prepared of each

weld

on

each connections.

These

weld sketches

were attached

to

number

86-0542.

Review of weld

sketches

showed

that at

least

one

weld

was

undersized

on

126

of

the

connections.

However, in all but

a

few cases,

the effect of the undersized

was offset

by oversized

in other

parts

of the

connection.

The

inspectors

examined

the

design

document

(design

change

notices,

field

change

requests,

and

modification

packages)

listed below which were issued to modify the

CB platform steel

connections.

The

inspectors

also

examined

the calculations

associated

with the modification

packages.

A

summary

of the,

design

documents

reviewed

by the inspectors

is listed below.

DCN 650-960,

issued

on March 19,

1986,

by

EBASCO to modify

connections

on

CB steel

platforms due to increases

in loads

from attachments

to platforms.

This

DCN involved the more

highly stressed

beams

which

had

been identified by

EBASCO

prior

to

the

time

the

structural

st'eel

design

was

transferred

from EBASCO to the

CP8 L onsite

HPES unit.

FCR AS-10481

issued

on

May 28,

1986 to revise details

on

DCN 650-960 for modifications to connections

on elevation

261

CB platform at azimuths 5'-9', 266',

30',

and

345'.

The

FCR

was

required

due

to field conditions

and

interferences.

FCR-AS-10546,

issued

on June

12,

1986, to revise details

on

DCN 650-960 for modification to connections

on elevation

286 at azimuths

330

and

345

.

The

FCR was required

due to

field conditions

and interferences.

Field Modification (FM)

C 6525,

issued

October 5,

1986, to

modify four connections

to increase

capacity of connections

to carry additional vertical loads.

10

FM

C

6526,

issued

October 4,

1986,

to

modify five

connections

due to torsional

loads.

FM

C

6527,

issued

October

5,

1986,

to

modify three

connections

due

to

loads

from the

main

steam

line whip

restraints.

One of these

connections

number 236-C3822N-188

had

been

undersized.

The

specified

weld size

(per

the

original design drawing) was 0.3125 inches.

The "as-built"

weld size

was 0.281 inches.

The required weld size,

due to

the increased

loads

was 0.345 inches.

FM

C 6529,

issued

October 4,

1986, to modify one connection

by installation of two 7/8

inch diameter

high

strength

bolts.

Modification was

due to increase

in loads acting

on

CB platform due to attachments.

FM

C

6531,

issued

October 7,

1986

to

modify five

connections

due to increased

loads

from attachments

to the

CB structural

steel

platforms.

The modifications involved

increasing

weld sizes

(beyond those originally specified),

or addition of high strength bolts to the connections

FM

C

6535,

issued

October 8,

1986,

to

modify

two

connections

due to torsional

loads

and to add seat plate to

one

connection

due

to

increased

vertical

loads

due

to

attachments.

The original designed

weld on this connection

was overstressed

due to new attachment

loads.

FM

C

6539,

issued

October 9,

1986,

to

modify

two'onnections

due to torsional

loads.

FM

C

6542,

issued

October 30,

1986,

to

modify

two

connections

due

to

increased

loads,

and

to

remove

the

modifications

installed

under

DCN 650-960

on

one

connection.

The

removal

consisted

of cutting the plates

attached

to the flanges

on Connection

Number 261-C3822N-575

so the connection

would behave

as pinned connection rather

than

a fixed connection after the plates

were installed.

The fixed connection

was undesirable

since

high stresses

were transferred

from the

beam into the

embed plate at this

connection.

FM

C

6543

modified

once

connection

due

to construction

problem

(nuts

were

not

tightened

on

bolts

and

w'ere

inaccessible

due to interferences

added

since the bolt were

installed).

Review of the calculations

and modifications listed above

showed

that the welds were modified (enlarged)

due to changes

in loads

during final platform steel

design 'verification

program,

not

because

of undersized

11

(3)

Findings

The

concern

was

not substantiated.

Although undersized

existed

on

some

connections,

modifications

to

the

welds/connections

were

not required

because

of the undersized

welds, but as

a result of additional

loads

from new attachments

~ to

the

structural

steel

platforms.

The

analysis

of

the

additional

attachment

loads

were part of a planned

systematic

design

program.

d.

Design of Welded Pipe Attachments

Concern

Problems

were identified with design of welded pipe attachment.

The criteria

used

to

design

the

attachments

(e.g.,

lugs,

trunnions,

etc)

was

inconsistent

and

was

not

properly

documented.

(2)

Di scussion

This is similar to the concern

expressed

in guality Check 9950.

An

evaluation

of welded

pipe

attachment

(WPA) calculations

disclosed

inconsistencies

in use of various coefficients (e.g.

Beta factors

and load coefficients).

To respond to the (}uality

Check,

licensee

engineers

conducted

a review of 125 pipe

hanger

designs

which included

a welded pipe attachment.

This review

disclosed that there

were inconsistencies

in selection

of load

coefficients

and

Beta

factors.

However,

all

calculations

reviewed were found to

have

used

acceptable

load coefficients,

although

in

some

cases

they

result

in

higher

calculated

stresses.

The review disclosed

that in

some calculations

very

conservative coefficients were

used

in calculating

load stresses

in the pipe wall at the

WPA, while in othe'r cases,

the correct

value listed in various Beta and/or load coefficient tables

was

used

in

the

calculation.

Use

of

the

conservative

load

coefficient

values

resulted

in calculation

of higher

local

stresses

in the pipe welds.

For'example,

since

pipe stress

is

-'~'--,.'.based

on the

load coefficient or Beta factor multiplied by the

applied

loads acting

on the pipe at

the

WPA,

use

of

a

Beta'actor

or load coefficient of 0.4 when the correct value

was 0.3

would result in a

33 percent

increase

is calculated

stresses

in

the pipe (0.4/0.3

= 1.33).

Therefore

even

though

an incorrect

factor

was

used

in

the

analysis,

its

use

had

no

safety

significant

since

the

errors

were

always

in the conservative

direction.

During review of the

125 calculation

packages,

minor errors were

found

in eight

supports

calculations.

These

errors

were

not

associated

with the

Beta

factor s or load coefficients.

The

12

,':l,icensee

issued

NCR 87-096

to document

and disposition

these

minor errors.

Correction

of the

problems

resulted

in minor

revisions

to the calculation

packages.

No work was required to

modify the hangers

affected

by the minor calculation errors.

In order to assure

that consistent

values for load coefficients

would

be

used

in future design of welded pipe attachments,

the

licensee

issued

NED

Design

Guide

No. DG-II.12,

Design

and

Analysis of Welded Pipe Attachments for the Harris Project.

The

inspector

reviewed this design

guide

and

found that it clearly

establishes, requirements

to

be

used

in calculation

of local

stresses

in pipe walls at the

WPA.

(3)

Finding

The concern

was substantiated

in that criteria used in design of

welded pipe attachments

was inconsistent.

However this problem

had

been previously identified and resolved

under the licensee's

guality Check Program.

Although inconsistencies

were identified

in

some calculations,

these

inconsistencies

did not result

in

nonconservative

piping designs.

The inconsistent

use of load

coefficients resulted

in more conservative

calculated

stresses

and thus did not affect any hardware.

Thus, this substantiated

concern

has

no safety significance.

Previously Identified Inspector

Follow-up Items

a.

(Closed) IFI 400/86-69-02,

Reinspection

of Shared Electrical Supports

for Deleted Conduit

During the September

4-11,

1986

NRC inspection,

a Region II inspector

examined

the

shared

conduit

support

problems

identified

on

86-0444.

The

NCR documented

discrepancies

identified by the licensee

during

walkdown

inspections

of non-nuclear

safety-related

(NNS)

conduits

which disclosed

that installation of

NNS conduit supports

that

share

supports

with

safety-related

raceway

had

not

been

inspected

for seismic attributes.

The cause

of this problem was due

to'insufficient instructions

in

CP&L Procedure

TP-51,

Inspection

of

Non-,Safety

Related Electrical

Raceways,

to assure

that installation

of,"'.NNS conduits that shared

supports with safety-related

(SR) conduit

would be inspected

to verify that seismic installation requirements

were completed

as noted

on Sheet

14 of Drawing number

CAR 2166-B-060.

'n order to correct this problem,

the licensee

amended

Revision

7 of

procedure

TP-51 with

a

procedure

change

notice

(PCN) to clearly

specify that

NNS conduit that shares

supports with SR conduit are to

be inspected

to assure installation conforms to seismic

requirements.

The

NNS conduits

listed

in

NCR 86-0444

were reinspected.

Conduit

found with unacceptable

seismic

installations

were

reworked

and

reinspected.

Some

of

the

rework

involved

removal

of

temporary

conduits that

had been installed for construction,

and other conduit

13

no longer required

due to design

changes

and thus were scheduled

to

be

removed.

These

conduits

are referred

to

as

"deleted

conduit".

During

the

September

4-11,

1986

inspection,

an additional

conduit

number

17129K,

on

a

shared

support

was

identified

which

lacked

inspection

records

to verify compliance

with seismic

attributes.

This problem was documented

on

NCR 86-0555.

To correct this problem,

conduit

17129K

was

inspected

for seismic

installation attributes.

The

installation

was

found

to

be

acceptable.

The

licensee's

corrective

actions to resolve

the shared

support

problems

included

a

detailed

walkdown program to identify all

NNS and "deleted"

conduits

which share

supports with SR conduit,

review of inspection

records

to

verify NNS conduit

on

shared

supports

had

been

inspected,

necessary

rework

to

correct

shared

support

problem

identified

during

inspection,

removal

of

"deleted"

conduit

where

required,

and

documentation

of inspection

of

NNS conduit

on

shared

supports

when

records

were

found to

be

incomplete.

In addition,

the

Regulatory

Guide

1.29

walkdown program,

discussed

in paragraph

6.d below,

was

another

program

conducted

in'dependent

of

the

resolution

of

86-0444

and

86-0555

to identify potentially inadequately

supported

non-safety

which could affect

SR equipment.

The

inspectors

examined

documentation

supporting closeout of NCR 86-0444

and

-0555

and the following memorandums

which

summarize

corrective

actions

to resolve

shared

support problems:

Memorandum

to

A. H. Rager

from

V. Cox/J. Martin

dated

September

22,

1986, Subject:

Deleted Conduits

Memorandum

to

M. Holveck

from

V. Cox,

dated

October 3,

1986,

Subject:

Response

to

Concern

Conduits Installed

on Shared

Supports

Memorandum

to

V..Cox

from J. Frantz,

dated

October

28,

1986, Subject: 'hared

Support Walkdown

Memorandum

to

File

HXDE-2XX-XXX-516 from

J. Martin /

M. Bodnar,

dated

November 24,

1987,

Subject:

Deleted

Conduit

Subsequent

to closeout

of

NCR 86-0444

and -0555,

NCR OP-86-0121

was

issued

on

October 24,

1986,

regarding

identification

of

two

addi.tional

conduits,

numbers

19567K

and

16231U,

which

had not

been

inspected

for seismic attributes

during original installation.

The

conduits

were

inspected

for

seismic

attributes

to

resolve

this

problem.

Conduit

16231U

was

acceptable

while conduit

19567K

was

found to

be

unacceptable

and

required

further evaluation.

After

evaluation

by HPES, this conduit was found to be acceptable.

Further

review

by the

licensee

resulted

in cancellation

of

NCR OP-86-0121

since

the

walkdown inspection

program

being

conducted

to identify

these

problems

was

in

progress.

The

inspector

concurs

with

cancellation

of this

NCR.

Based

on review of the licensee's

program

for reinspection

of shared

supports,

IFI 400/86-69-02 is closed.

(Closed) IFI 86-77-04,

Resolution of Cable Tray Riser Design Concerns

An engineer,

designated

Individual

E in Inspection

Report

Number

50-400/86-77,

involved with the analysis of the containment

building

cable tray riser structures

re'signed

his position at the Harris site

to accept

other

employment.

Since

he

had only partially completed

the cable tray riser analysis,

Individual

E prepared

a list of his

pending

concerns

regarding

the analysis

and

submitted this list to

his supervisor.

The licensee

addressed

these

concerns

in calculation

number

LV-66,

Cable

Tr ay

Ri ser

Frame

(Final

Verification) .

The

-inspectors

reviewed

calculation

LV-66 to determine if the

design

engineer's

concerns

were

resolved.

During

review

of

this

calculation,

the inspectors

noted that the change

in temperature

(hT)

used

in calculation of thermal

stress

in the structure

was

assumed

to

be

60OF

not

the

value

of

148

F

used

in the

CB structural

steel

platform analysis.

This error appeared

on page

67 of the calculation

and

was

used

in recalculation

of thermal

stresses

after

use of the

correct

bT value of

148

F resulted

in

an interaction

equation

of

2.47, which considerably

exceeds

the acceptable

value of 1.0.

The change

in the

hT value from 148'o 60'F constitutes

a change

in

a specified design input, specifically, the accident

temperature

used

to

calculate

thermal

stresses.

This

change

was

not

properly

identified, approved,

documented

or controlled in accordance

with the

licensee's

approved

QA

program

or

10 CFR 50,

Appendix B,

Criterion III. This was identified to the licensee

as Violation Item

400/87-41-02,

Uncontrolled Change to Design Input.

After the inspector identified the

problem regarding

the

change

in

design

input for

hT,

licensee

engineers

conducted

a

review of

calculation

LV-66.

This

review disclosed

that incorrect allowable

stresses

had

been

used

in

some

portions

of the calculations

to

qualify

some

members for the operating

base

earthquake

design

loads.

These

problems,

the

incorrect

hT

and

the

incorrect

allowable

stresses,

were

documented

on

NCR 87-51.

During resolution of the

NCR,

which

involved

an

extensive

revision

to calculate

LV-66,

licensee

engineers

discovered

some errors in geometry in the computer

model of the riser structure

analyzed

in the original calculation

LV,.-'66.

As

a result,

licensee

engineers

conducted

an extensive

design

re;.."evaluation of the riser structure.

This re-evaluation

was still

ongoing at the conclusion of the inspection.

Based

on review of calculation

LV-66, the inspectors

determined that

Individual E's concerns

regarding

the cable tray riser frame analysis

were

addressed

in calculation

LV-66, although

the errors

discussed

above will result

in revision of the calculation.

A

summary

of

Individual E's concerns

and their resolutions

follows below:

4

15

Concern

No frequency calculations

for attaching

conduit

and instrument

supports

were performed.

Resolution

(2)

A frequency analysis

was performed

and considered

in analysis of

attachment

of conduit

and

instrument

supports

to the riser

frame.

The attachments

were

checked

for

a

reduced

frequency

determined

from

an

updated

static

analysis,

based

on actual

column properties.

'Concern

The

column

properties

input

in

the

EBASCO

analysis

were

incorrect.

The

moment

of inertia

( I) input in the

EBASCO

analysis

assumed

the

2

C9 x 15

members

were

adjacent

to

each

other

and

behaved

as

a

beam without consideration

of actual

riser configuration

(Channels

spaced

approximately

15

inches

center-center

with cross-braces

consisting

of horizontal tie

plates

connecting

the

channels).

The

same

condition

also

applies to the internal bracing (built-up 4 x 4 x 3/8 angles).

Resolution

The actual "as-built" riser configuration of the riser structure

(channels

connected

by the plates)

was

modeled

in the analysis

and

an

equivalent

moment of inertia

was

computed 'or

these

members.

The equivalent

moment of inertia

was

used

in

the

analysis.

The internal

bracing

was

also

analyzed

using

the

correct

moment of inertia.

(3)

Concern

There

were

no calculations

performed to determine

stresses

in