ML18038B592
| ML18038B592 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/15/1995 |
| From: | Casto C, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18038B590 | List: |
| References | |
| 50-259-95-57, 50-260-95-57, 50-296-95-57, NUDOCS 9512270144 | |
| Download: ML18038B592 (26) | |
See also: IR 05000259/1995057
Text
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UNITED STATES
NUCLEAR REGULATORY'COMMISSION
REGION II
101 MARIETTASTREET, N.W., SUITE 2900
ATLANTA,GEORGIA 303234199
Report Nos.:
50-259/95-57,
50-260/95-57,
and 50-296/95-57
Licensee:
Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga,
TN
37402-2801
Docket Nos.:
50-259,
50-260
and 50-296
License Nos.:
and
Facility Name:
Browns Ferry Nuclear
Power Station Units 1,, 2,
and
3
Inspection
Conducted:
October 2-6, October 17-20,
November 1-3,
November 6,
and
November 13-17,
1995
Inspector:
J. J.
Lenahan
Accompanying Personnel:
C.
Chou
(Hove
Approved by:
C. A.
asto,
Chief
Engineering
Branch
Division of Reactor Safety
/z rs/PW
Date Signed
1-3,
and November 6,
1995)
l5
5'
Date Signed
SUMMARY
Scope:
This routine,
announced
inspection
was conducted
in the areas of small
pipe supports,
instrument tubing and supports,
seismic II/I issues,
Unit 3
.
startup
issues,
and licensee
action
on previous inspection findings.
On
November 6,
1995,
a meeting
was held in
NRC Headquarters,
Rockville, Maryland,
between
NRC and the l.icensee to discuss
design details regarding structural
steel
beams
and coverplates.
Results:
In the areas
inspected,
deviations
were not identified.
A violation was identified for failure to implemen't installation of small
piping, and instrument supports,
and structural
steel
in accordance
with design
drawing requirements,
paragraphs
2.1-, 2.4,
and 4.4.
A weakness
was identified
for failure to perform
a generic review how an issue identified on Unit 3
affected'he
Unit 2 operating plant - paragraph
3.0.,
Enclosure
2
9512270144
951215
ADOCK 05000259
8
0
4l
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REPORT DETAIL
1.0
Persons
Contacted
Licensee
Employees
¹J.
- R.
J.
¹*D.
- G
¹*L.
- G
- p
"H.
- S
¹*J.
Corey,
Manager,
Radiological
Control
and Chemistry
Gilbert, Operations
Support Supervisor
Glass,
Supervisory Civil Engineer,
Nuclear Engineering
(NE)
Housley, Site Licensee
Engineer
Little, Operations
Superintendent
Madison, Unit 3 Civil Engineering Supervisor
Preston,
Plant Manager
Salas,
Licensing Manager
Williams, Engineering
and Materials Manager,
NE
Wetzel, Acting Compliance
Manager
Valante, Unit 3 Lead Civil Engineer,
NE
2.0
2.1
Other licensee
employees
contacted
during this inspection
included
craftsmen,
engineers,
technicians,
and administrative personnel.
Other Organizations
P. Koski, Structural
Engineer,
Bechtel
V. Kapoor,
Browns Ferry Civil Design Unit Manager,
Bechtel
NRC Resident
Inspectors
¹L. Wert, Senior Resident
Inspector
R. Husser,
Resident
Inspector
- J. Hunday,
Resident
Inspector
- Attended October 20,
1995 exit interview
¹Attended
November
16,
1995 exit interview
Followup on Unit 3 Restart
Issues
Cable Tray and Conduit Support
guestions
were raised
by
NRC and through the employee
concerns
program
regarding
seismic qualification of cable tray and conduit supports.
The
resolution of this issue
can
be subdivided into two categories:
new
conduit
and cable tray supports,
and evaluation of existing cable tray
supports.
New supports
are those installed since
1986.
These
supports
are designed
in accordance
with the licensee's
revised seismic design
criteria and were installed under the licensee's
current quality
assurance
program requirements.
Existing supports
were those installed
prior to 1986, primarily during original plant construction.
Inspection
and acceptance
of existing conduit
and cable tray supports
was performed
under the Generic
Implementing Procedures
(GIP) for Seismic Verification
Enclosure
2
i%I
,<Qi
of Nuclear Plant Equipment.
The GIP was issued
by the Seismic
Qualification Utility Group
(SQUG) in response
to
NRC Unresolved Safety
Issue
A-46 (USI A-46), Seismic
Adequacy of Mechanical
and Electrical
Equipment in Operating Plants.
During the inspections
documented
in NRC Inspection
Report
numbers
50-
259,260,296/95-15
and 95-52, the inspector
examined
new conduit
and
cable tray supports,
and the implementation of the A-46 program for
evaluation of existing supports.
These
programs
were found to be
acceptable
in all areas,
except the drywell, where the licensee's
A-46
walkdown program
was still in progress.
The licensee
recently completed
the A-46 program for conduit
and cable tray supports
in the drywell.
During the current inspection,
the inspector
performed
a walkdown
inspection in the Unit 3 drywell to assess
the effectiveness
of the
licensee's
A-46 cable tray and conduit inspection
program.
The
inspector
examined existing conduit
and cable tray supports
on drywell
elevations
563,
584,
604
and 616.
With the exception of a missing
conduit clamp,
no deficiencies
were identified regarding the cable tray
and conduit supports.
Work was in progress
on dressing
cables
and
installing cable tray covers
on cable trays in the electrical
area
between elevation
575 and 584.
During the walkdown the
following deficiency was identified by the inspector:
A pipe clamp was
found to be missing
on
a
J inch diameter control air pipe support at
elevation
605, Azimuth 195 degrees,
in the Unit 3 drywell.
Drawing
number 3-478600-2033,
Revision I, specifies installation of a clamp
on
this support,
designated
support
sequence
number Ol, to ensure that
a
three directional restraint of the pipe was provided.
The failure to
install the clamp in accordance
with the design drawing requirements
was
identified to the licensee
as Violation item 296/95-57-01,
Failure to
Install Modifications in Accordance with Design Drawing Requirements.
During the walkdown inspection,
the inspector also identified other
minor items
such
as
a missing clamp
on
a non-safety related
pipe
support,
loose grating
on
a platform under the reactor vessel,
an
apparent
leak on
CRD module 50-23,
and
some
housekeeping
items.
The
loose. grating is discussed
in paragraph
3.0,
below.
The inspector
reviewed the results of the vessel
hydrostatic test
and verified that
the leak from the
CRD module met the hydrostatic test
acceptance
criteria.
The leak from CRD Module 50-23 .was observed
to be
3 drops per
minute .(dpm) during the hydrostatic test,
which is well within the 30
dpm hydrostatic test acceptance
criteria.
The remaining
items were
addressed
by the licensee
on work requests.
The inspector
concluded that the licensee's
A-46 cable tray and conduit
walkdown inspection
and evaluation
program meets
NRC requirements
and is
acceptable
for Unit 3 restart.
Enclosure
2
ik
]pi
2.2
Seismic Class II/I - Water Spray
This issue
involved the effects of water spray from potential failure of
non-safety related piping on safety related
equipment.
The Unit 3
program
implemented to evaluate this issue
was similar to the
one
implemented for Unit 2 prior to Unit 2 restart.
The inspector
reviewed
EgE Report
No. 51001.23-R-001,
Revision 0, dated
September
30,
1993,
Evaluation of Seismic
Induced II/I Spray Hazards at Browns Ferry Unit 3.
This report described
the program which included inplant screening to
identify potential
hazards,
evaluation of outliers,
recommended
plant
modifications,
and configuration control guidelines.
The screening
resulted
in identification of ill potential outliers.
The inspector
reviewed Calculation
number CD-(3999-931257,
Evaluation of Seismic-
Induced II/I Spray Hazards,
which was completed to evaluate
the
outliers.
Ninety-five of the outliers were resolved
by evaluation,
while the remaining
16 required modification.
The modifications were
implemented
under
DCN Number
W 22556A.
The inspector reviewed the
design drawings in the
DCN package
and verified that the
16 outliers
which required modifications were covered
under the
DCN.
The inspector
examined
support
number 3-47B491-3000
which was installed
under the
DCN
to correct
one of the outliners.
The remaining modifications were
inaccessible.
2.3
2.4
The inspector concluded that the licensee's
program complied with their
commentments
and
was acceptable
for Unit 3 restart.
Instrument Tubing
This issue
concerned
seismic re-analysis
of instrument tubing.
The
reanalysis
resulted
in modifications to a large number of instrument
tubing supports.
The licensee
combined the instrument tubing program
with the small
bore piping program.
Inspection of instrument tubing is
covered
under paragraph
2.4,
below.
Small
Bore Piping
This issue
concerned
the seismic qualification of existing class
1 small
bore piping and instrument tubing and resolution of deficiencies
identified in the small
bore piping/instrument tubing.
The small
piping in this program included all Class
I piping systems
less
than two
inches in diameter which had not been rigorously analyzed.
The
licensee's
commitments for resolution of this issue
are specified in a
letter to NRC, dated
February
27,
1991, Subject:
Brown Ferry Nuclear
Plant - Action Plan to Disposition Concerns
Related to Units
1
and
3
Small
Bore Piping.
The program included performance of walkdown inspections
to determine
as-built conditions for small
bore piping/tubing; performance of
rigorous analysis;
preparation of design
changes
to repair identified
Enclosure
2
deficiencies
and install
new supports;
installation of the design
changes;
and performance of a confirmatory analysis
on ten percent of
the small
bore piping/instrument tubing
and associated
supports.
The inspector
reviewed Engineering Attribute Walkdown Instruction
BC-
012,
Walkdown Instructions for Seismic Class
I Small
Bore Piping, Tubing
and Associated
Supports.
This instruction established
the requirements
and criteria for performance of walkdown inspections of small
piping/tubing in order to identify discrepancies
and obtain as-built
data for the rigorous analysis
program.
The inspector
performed
a walkdown inspection
and examined
randomly
selected
supports
on various
systems
to determine if modifications
identified from the rigorous analysis
program were completed in
accordance
with design requirements.
Acceptance criteria utilized by
the inspector
were critical specific in Hodification and Addition
Instruction HAI-4.2A, Piping/Tubing Supports,
Revision
19.
The
following small
bore piping and instrumentation
tubing supports
were
inspected:
RHR System,
support
numbers
3-47B452-3022,
3069 through 3074,
3076,
3078,
3138 through 3142,
3178,
3180,
and 3181.
Core spray system,
support
numbers
3-47B458 -731,
-735, -740,-
741,
-745 through -749,
-757 through -764,
and -768.-
Control air system,
support
numbers
3-47B600 - 1433 through
1436,
-1439 through
1442, support
sequence
numbers
01 through
12 on
drawing numbers
3-47B600 -1432,
and
sequence
numbers
Ol through
05
and
15 -
17 on drawing number 3-47B600 -1438,
Reactor Mater Cleanup
(RWCU) system,
support
number 3-47B456 -258
RCIC system,
support
number 3-47B456
-258
Reactor water recirculation system,
support
numbers
3-47B465 -587,
-589,
2004 through
2006,
2014,
and 2015.
CAD system,
support
numbers
3-47B461
-772 through
774
Containment inerting system,
support
numbers
3-47B650
-25 and -26.
The following discrepancies
were identified during the walkdown
inspection:
Enclosure
2
0
<5
<Qi
An oversized
hole was not repaired in the control air system
support,
designated
as support
sequence
03
on drawing number 3-
478600
-1432, Revision
1.
Note
12 on the drawing required that
the oversized
hole
be repaired.
A quarter inch thick plate with no gap
was observed to have
been
installed
between
the clamp halves
on the anchor/support
for
control air system support
number 3-478600-1435.
Note
2 on
drawing number B-47B600-1435,
Revision 0, requires
the clamp
halves to be installed with a gap greater
than zero inches.
In addition the resident
inspector identified the following discrepancy
while inspecting the reactor vessel
level backfill modification on
instrument
panel
number 3-LPNL-925-674 and -675.
Note
4 on drawing
number 3-47B600-2512-2,
Revision 01, requires that tubing clamps
be
installed to ensure that the maximum unsupported
tubing span with
concentrated
weights (i.e.,
unsupported
valves
and flow elements)
does
not exceed
24 inches.
Detail
B on drawing number 3-47B600-2512-1,
Revision 0, requires that
a structural
support
be provided
on the panel
for instrument tubing and flow elements
in accordance
with Detail J,
drawing 3-47E600-808.
Contrary to this requirement,
tubing clamps were
not installed to secure
the tubing/flow element to the Detail J support.
The unsupported
tubing span
exceeded
24 inches.
These three discrepancies
were identified to the licensee
as additional
examples of violation item 296/95
57 01.
The licensee
issued
problem evaluations
reports to document
and
disposition the above problems.
Prior to the end of the current
inspection,
the inspector verified that the above
items
had
been
repaired prior to restart of Unit 3.
The inspector also identified several
other minor discrepancies
which
were not considered significant enough to identify as additional
violation examples.
These
included minor drafting errors
on two
drawings which were documented
on
PER number
BFPER951556,
and
a minor
error in welding as
RHR support which was documented
on
PER number
BRPER951541.
The drawings
were corrected.
In addition,
RHR instrument
tubing attached
to support
numbers
3-47B452-3140,
3141,
and 3142 were
found to be out of tolerance.
This problem was documented
on
PER number
BFPER951426.
The pipe clamps
were adjusted
on the supports
under work
order C294977.
The inspector reinspected
these
supports
and verified
that the clamps
had
been adjusted
so instrument tubing was installed in
accordance
with design drawing requirements.
The licensee's
Unit 3 small
bore piping/instrument tubing program
had
also
been previously inspected
during inspections
documented
in
NRC
Inspection
Report
numbers
50-259,260,296/92-19
and 95-03.
Based
on the
Enclosure
2
0
~$
3.0
results of the current inspection,
the inspector
concluded that the
licensee's
Unit 3 small
bore piping/instrument tubing program
was
implemented
in accordance
with their commitments to
NRC and is
acceptable
for restart of Unit 3.
Within the areas
inspected,
violations or deviations
were not
identified.
Walkdown of Structural
Steel
Platforms
The inspector walked
down structural
steel
platforms
and examined the
attachment of grating to the platform structural
steel
members.
The
inspector also
examined
methods
used to anchor equipment to the platform
steel
and/or grating.
During the walkdown inspection,
the inspector
determined that the grating
had not been attached
to the platform steel
in accordance
with the details specified
on the design drawings.
As
discussed
in paragraph
2.1 above,
the inspector identified loose grating
on
a platform located
under the Unit 3 reactor vessel.
Since the
drywell closure
was not completed,
and the licensee
had
an open work
order,
number 95-17260-00,
to complete this work, this problem was not
identified as
a violation.
The inspector
re-examined this area prior to
the
end of the inspection
and verified that the grating was properly
attached
to the structural
steel prior to drywell closeout.
The
inspector also identified areas
in the Unit 3 corner
rooms where the
grating
had not been positively attached
to the structural
steel
as
required
by paragraph
4.2.5. 1A of TVA General Specification
G-89,
Structural Steel,
Revision 3,
and
TVA drawing number 0-48E915,
Revision
2, Hiscellaneous
Steel
Floor Grating
and Railing.
Since work in this
area
had not yet been
completed, this problem also
was not identified as
a violation.
The licensee
issued
work request
C294942 to complete this
work.
The inspector re-examined
the areas prior to Unit 3 startup
and
verified that the grating was properly attached
to the structural steel.
During the walkdowns the inspector
noted
some miscellaneous
equipment
and small structural
steel
platforms were anchored to the platform
grating.
The licensee
performed
a design evaluation of the worst case
small steel
platform and checked for stability against overturning
and
sliding.
The inspector
reviewed the design calculations
which showed
the anchorage
of the platform was adequate.
The inspector also
determined that the equipment (electrical
conduits
and junction boxes,
and small
bore piping) was adequately
anchored to meet design loading
conditions.
Approximately two weeks after the inspector
had identified the issues
regarding
loose grating in Unit 3, the resident
inspector identified
a
similar problem in Unit 2.
The licensee
issued
Problem Evaluation
Report
BFPER951655 to document
and disposition this problem.
The
licensee
performed
an operability review and determined that the loose
sections of grating would not affect operability of any safety-related
Enclosure
2
IN
iO
<Qi
equipment.
However, the inspector questioned
why licensee
personnel
had
not identified the problem
on Unit 2 after the problem in Unit 3 had
been previously identified.
Since
a
PER had not been
issued for the
Unit 3 problem,
the extent of condition review for affect on other Units
had not been performed.
The inspector identified the failure to perform
a generic review of the Unit 3 loose grating issue
and its possible
affect on Unit 2 as
a weakness
to licensee
management.
Within the areas
inspected,
violations or deviations
were not
identified.
4.0
4.1
Action on Previous
Inspection
Findings
(Closed) Violation Item 296/95-15-01,
Failure to Complete Modifications
in Accordance with Work Plan
and Drawing Requirements
The licensee
responded
to this violation in a letter dated
June
20,
1995, Subject:
NRC Inspection
Report 50-259,260,296/95-15
- Reply to
The licensee attributed the cause of the two
examples
in the violation to personnel
error during installation of the
modifications.
Example
1 involved failure to upgrade
a weld on
a beam
as required
by the design drawings.
Problem Evaluation Report
(PER)
number
BFPER 950293
was issued to documented
and disposition this
problem.
Example
2 involved failure of quality control personnel
to
inspect
a conduit support.
PER number
BFPER 950273
was issued to
document
and disposition this problem.
The licensee
issued
a work order
and upgraded
the welds
on the
beam in
example
1 to the size specified
on the design drawings.
The inspector
examined
the modified welds
and verified that they complied with the
design requirements.
Additional corrective action included counseling
of civil/structural site engineering
personnel
regarding this problem,
and the necessity
to provide adequate
and specific location information
on design drawings.
The licensee
issued
another work order to document
installation
and inspection of the support in example
2 in accordance.
with their procedural
requirements.
The support did not require
any
additional modification.
The inspector
reviewed the completed
inspection records
and verified that the inspections
had
been
completed
in accordance
with the licensee's
procedures.
Additional corrective
actions
include counseling
and training of all field engineers
to
emphasize
the importance of proper closeout of work plans,
and
a
statistical
sample to review documentation for conduit support
installations.
Enclosure
2
0
il
<Qi
8
(Closed)
Unresolved
Item 260/95-52-01,
SDV System Inspection
Following
Reactor
4.3
This unresolved
item concerned
implementation of the licensee's
commitment to inspect the scram discharge
volume system after the first
scram from rated temperature
and pressure
following a refueling outage.
The commitment
was
made in response
to Generic Safety Issue
(GSI) 40-
Safety Concerns
Associated with Pipe Breaks in
Review
of operations
data (control
room logs
and computer printouts)
showed
that the
SDV system
was walked down and inspected within 30 minutes
on
August 9,
1991, following the first scram which occurred
from full
temperature
and pressure
following the refueling outage (first cycle
following Unit 2 restart).
The data
was not available for review for
June 3,
1993 scram
(second
cycle following Unit 2 restart).
For the
December
2,
1994 scram (third cycle following Unit 2 restart),
the
control
room logs indicate that the
SDV system walkdown was not
completed within the 30 minutes time period
as per the licensee's
commitment.
However, further review disclosed that the
SDV system
was
subjected
to
a hydrostatic test prior to startup after the last Unit 2
refueling outage.
The hydrostatic test satisfies that licensee's
commitments for SDV system
walkdown inspection.
The licensee
revised
their commitment for the periodic walkdown inspection of the
SDV system
in a letter to
NRC dated
November 21,
1995, Subject:
Browns Ferry
Nuclear Plant - Revisions to Commitments
Concerning the Utilization of
Security Personnel
for Notification of Irradiated Fuel'amage,
Replacement
of the Process
Computer,
and Periodic Walkdown of the Scram
Discharge
Volume Piping System.
The revised
commitment requires
a
system leakage
inspection
once per refueling outage during the reactor
vessel
hydrostatic test.
(Closed) Violation Item 296/95-52-02,
Failure to Construct
Cable Tray
Support in Accordance with Design Requirements
The licensee
responded
to this violation in a letter dated
November 9,
1995, Subject:
NRC Inspection
Report 50-259,260,296/95-52
- Reply to
The licensee attributed the cause of the violation
to be personnel
error due to a poorly coordinated field design
change.
The licensee
issued
PER Number
BF
PER 95 1125 to document
and
disposition this problem.
A work order was issued to complete the two
missing flare bevel welds.
The inspector re-inspected
the support
and
verified the welds were completed
in accordance
with design
requirements.
The inspector also reviewed the quality control
inspection records
which documented
inspection of the welds.
The
licensee
issued
a memorandum to field engineering
personnel
which
summarized
the problem and reminded personnel
of their responsibility to
properly coordinate
design
changes.
A statistical
sample of other
new
cable tray supports
installed in Unit 3 were walked down to verify that
all welds were completed
as required
by design
documents.
No additional
problems (i.e., missing welds) were identified.
One other issue
was
Enclosure
2
<
lgi
identified, which had also
been included
on
BF PER 95 1125,
concerned
errors
made
when tagging cable tray supports with identification
numbers.
A large
number of errors in cable tray tag numbers
were
identified.
However,
these errors
had
no safety significance.
The
licensee
issued
work orders to correct the tag numbers.
4.4
(Closed)
Inspector
Followup Item (IFI) 260,296/95-41-02,
Platform Steel
gualification
This IFI was identified due to questions
regarding calculations
used to
qualify reinforcement
elements
on the Unit 3 drywell platform structural
steel
beams.
The problems
involved effective areas,
cut-off points,
critical load cases,
and selection of the critical sections to be
checked.
Problems identified were
as follows:
In Calculation CD-(3303-920114,
Revision 6, the interaction ratio
was not checked for the applied stress
against
the allowable
stress for Operating
Base
(OBE) case for Member
5
(8WF17) from 0.7L to 0.9L to qualify the member.
The section
modulus
used for this member in Revision
6 of the calculation
was
calculated
using conservative
assumptions.
The section
modulus
for the member
was recalculated
using the actual
beam properties
which were used to check the applied stresses
in the member.
The
revised calculations
showed the applied stresses
were less
than
allowable stress
values.
Calculation CD-(3303-930573,
Revision 6, qualified composite
section
k2-k2 on Drawings 3-48E453-2;
Revision
1
(DCA W17767-003,
Revision 0).
This composite section
was
a wide flange
10WF21
beam
reinforced
by a T-section
(WT Bx15.5) attached
(welded) to the web
in the horizontal direction.
The problem concerned
the licensee's
failure to check this composite section
using the full effective
area to reduce the allowable stress for the ineffective
beam
elements,
or to reduce the ineffective area to develop the full
allowable stress
per AISC Code requirements.
The licensee
revised
the calculation to check the unstiffened lengths
using all the
elements
in the composite section to meet the code requirements.
All the elements
were acceptable
and could
be used for the fully
effective areas.
Therefore, this problem was resolved.
Calculation CD-(3303-920114,
Revision 6, qualified
a composite
beam which consists of a wide flange
10WF27 with a ~/i inch thick
side plate welded
between
the
beam flanges
on one side of the
10WF27,
and .a Bx6xj inch tube steel
section
welded to the side
plate using stitch welds.
During review of this calculation,
the
Enclosure
2
jgi
10
inspectors
noted that the licensee
had not evaluated
the effect of
transfer of stresses
from the tube steel
into the plate
and
ques i
uestioned
the licensee
regarding behavior of the composite
section for integration of stresses.
The licensee
fabricated
and tested
a test
beam which was similar
to the composite
beam.
The test
beam was fabricated
from a
W12x27, % inch thick plate
and Sx8xj inch tube steel
section.
The
purpose of the test
was to demonstrate
that
beam theory was
applicable for the composite
member.
The test
beam
was clamped in
test frames at its two ends
(simple supported)
and loaded using
a
hydraulic jack.
The load was applied to the flange of 'the W12x27,
to duplicate field conditions.
When the test load reached
60,900
pounds,
the anchorage for the test
frame failed, ending the test.
The test
beam was examined
by licensee
engineers
during and after
the test.
There were
no indications of weld cracks,
or damage to
or distortation of the
beam.
The maximum stress
at the extreme
edge of the
WF beam under the test load was 30,600 psi which was
below the allowable value of 32,400 psi.
The inspectors
examined
the. test
beam
and verified that it was
constructed
to be representative
of the actual as-built composite
member existing in the drywell. The inspectors verified the member
sizes
used in the test
beam were
as stated
by the licensee
and
that the method
used
(4 inch stitch welds at
12 inch centers)
to
fabricate the composite section duplicated field conditions.
The
inspectors
also verified that the test
beam
had not been
damaged
during the test.
There were
no indications of distortation or
weld cracking.
The inspectors
concurred with the licensee's
conclusions that the test demonstrated
that
beam theory was
applicable for the composite
member.
The licensee
revised the design calculations to incorporate
shear
flow theory to calculate
combined
bending
and shear stresses
acting at various points in the composite
cross section.
The
calculated
stresses
were less than allowable stress
values.
The
licensee's
design methodology were also discussed
during the
meeting with the Office of Nuclear Reactor Regulation
(NRR)
discussed
in paragraph
5.0,
below.
During the November
6 meeting,
NRR personnel
requested
that the
licensee verify that the composite
member
was constructed
accordance
with design drawing requirements.
When licensee
engineers
performed
a walkdown inspection of the
member
on
November 8,
1995, they discovered that the welds connecting the ~/i
inch thick plate to the
12
WF27 beam flanges
were not the
221 inch
continuous
length at each
end of the
member
as specified
on the
design drawings.
The actual
welds were only approximately
12
inches long.
The licensee
documented this problem on
PER number
Enclosure
2
.4
il
<p
<Ql
5.0
BF PER 951704.
A work order
was initiated to correct the welds.
The inspectors
examined the member
and verified it had
been
repaired
{welded) in accordance
with the details
on the design
drawing,
DCA WI7538-086, Revision 5.
The licensee
revised the
drawing and issued
Revision
6 to clarify the details for
fabrication of the composite
member.
The inspectors
discussed
the
extent of Condition evaluation of this
PER with licensee
engineers.
In evaluating the extent of condition, the licensee
inspected
the
6 other subframes
which had
been modified and
reviewed the
DCAs issued for structural
steel platform steel
modifications.
No other weld problems
were identified.
The
licensee
also reviewed the design drawings for the similar
subframe in Unit 2.
The review disclosed that different design
details
were
used to reinforce the member.
Failure to fabricate
the member in accordance
with design requirements
was identified
to the licensee
as another
example of Violation item 296/95-57-01.
This example is not considered
licensee identified since the
licensee
discovered
the incorrect welds
as
a result of questions
raised
by
NRC personnel.
Meeting
Summary
On November 6,
1995,
a meeting
was held in the
NRC Headquarters
Office in Rockville, Maryland to discuss
the design methodology
used in design of cover plates
on
beam sections.
The attendees
at
the meeting were
as follows:
G. Bagchi, Chief, Office of Nuclear Reactor Regulation, Civil
Engineering
and Geosciences
Branch
(NRR/ECGB)
D. Jeng,
Section Chief,
NRR/ECGB
J. Williams, Project Manager,
R.
Chou,
Reactor
Inspector,
Region II
J.
Lenahan,
Civil Engineer,
Region II
P. Salas,
Browns Ferry Licensing Manager,
J. Valante,
Browns Ferry Lead Civil Engineer,
G. Thomas,
Senior Structural
Engineer,
Bechtel
Corporate Offices
P. Koski, Structural
Engineer,
Browns Ferry Project,
Bechtel
Topics covered
in the meeting included discussions
of methods
used
to calculate stress
in beam composite sections,
weld design,
and
selection of critical sections for analysis.
The licensee
also
discussed
the results of a load test they performed
on
a mock up
model of a composite section to demonstrate
that the composite
section
behaved
as
a beam.
The test results
were acceptable.
personnel
requested
that the inspectors
perform an inspection to
confirm that the test
beam was fabricated to duplicate field
conditions,
and that the as-built
beam
was fabricated in
accordance
with design drawing requirements.
The results of the
inspection
are discussed
in paragraph
4.4,
above.
Enclosure
2
il
~
i
12
6.0
.Exit Interview
The inspection
scope
and results
were summarized
on October 20,
1995
and
November
16,
1995, with those
persons
indicated in paragraph
1.
The
inspector described
the areas
inspected,and
discussed
in detail the
inspection results listed below.
Proprietary information is not
contained
in this report.
D'issenting
comments
were not received
from
the licensee.
Violation Item 296/95-57-01,
Failure to Install Modifications in
Accordance with Design Drawing Requirements
Enclosure
2
'1
0
<Qi