ML18038B453
| ML18038B453 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/11/1995 |
| From: | Blake J, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18038B451 | List: |
| References | |
| REF-GTECI-040, REF-GTECI-NI, TASK-040, TASK-40, TASK-OR 50-259-95-52, 50-260-95-52, 50-296-95-52, NUDOCS 9510170396 | |
| Download: ML18038B453 (28) | |
See also: IR 05000259/1995052
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIE%IA STREET, N.W., SUITE 2900
ATLANTA,GEORGIA 30323.0109
Report Nos.:
50-259/95-52,
50-260/95-52,
and 50-296/95-52
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
Facili.ty Name:
Browns Ferry Nuclear, Power Station Units 1, 2,
and
3
Inspection .Conducted:
August 21-25,
and September
11-14,
1995
Inspector:
J. J.
Lenahan
Date Signed
Approved by:
Jerome J.
ake
Chief
Materi
.and
P ocesses
Section
Engineering
ranch
Division of Reactor Safety
/0 /i
5'5
Date Signed
SUMMARY
Scope:
-This special,
announced
inspection
was conducted
in the areas of Generic
Safety
Issue
(GSI) 40,
Pipe Breaks in
modifications of
heating, ventilation
and air conditioning
(HVAC) supports;
large bore piping
and supports;
cable tray and conduit support issues;
long term torus
integrity; platform thermal
growth; moderate
energy line break; control rod
drive
(CRD) piping support modifications; Unit 3 startup issues;
and licensee
action
on previous inspection findings.
Results:
In the areas
inspected,
violations or deviations
were not identified.
One violation was identified for failure to implement installation of a cable
tray support in accordance
with design drawing requirements
- paragraph
3.7. 1.
An unresolved
item was identified regarding
performance of walkdown
Enclosure
2
95ao<70SVS
95<0>2
,PDR
ADOCK 05000259
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2
inspections of the scram discharge
volume system. piping after the first scram
from full power,
and operating temperature,
following a refueling outage-
paragraph
2.
A weakness
was identified regarding errors
made during the drawing rollup
process-paragraph
3.7. 1 and 3.7.2.
The following issues
are resolved for Unit 3 restart:
GSI 40;
HVAC support
modifications; large bore piping and supports;
platform thermal
growth,
moderate
energy line break,
CRD pipe support
frame modifications,
and long
term torus integrity.
REPORT
DETAILS
Persons
Contacted
Licensee
Employees
- T
R.
¹*J.
¹*D.
- J
- J
¹*L.
¹D.
¹P.
- J
¹*H.
¹*S.
¹C.
Abney, Unit 3 Recovery
Manager
Gilbert, Operations
Procedure
Group Supervisor
Glass,
Acting Lead Civil Engineer
Housley, Site Licensing Engineer
Johnson,
Site guality Manager
Haddox, Maintenance/Modification
Manager
Madison, Unit 3 Civil Engineering Supervisor
Hatherly, Operations
Supervisor
Salas,
Licensing Manager
Valente, Unit 3 Engineering
Manager
Williams, Engineering
and Materials
Manager
Wetzel, Acting Compliance
Manager
Woods, Unit 3 Recovery Field Engineering
Other licensee
employees
contacted
during this inspection
included
craftsmen,
engineers,
technicians,
and administrative personnel.
NRC Resident
Inspectors
- L. Wert, Senior Resident
Inspector
¹R. Husser,
Resident
Inspector
J.
Hunday,
Resident
Inspector
- Attended
August
2S exit interview
¹Attended
September
14 exit interview
Generic Safety Issue
(GSI)
40 - Safety Concerns
Associated with Pipe
Breaks in The
On January
3,
1986,
NRC issued
Generic Letter 86-01 which accepted
the
BWR owners'roup
response,
BWROG-8420 to resolve
GSI 40.
BWROG-8420
proposed that leak detection
on the Scram Discharge
Volume
(SDV) system
be performed
once per refueling cycle by a walkdown within 30 minutes of
scram reset after the first scram from full power and temperature
following each refueling outage.
In an
NRC letter to TVA, dated
November 7,
1990,
NRC listed
an outstanding
issue regarding
GSI 40.
The
outstanding
issue
concerned
the fact that the licensee's
emergency
response
procedures
for Units
1 and
3 did not address
the visual
inspection of the
SDV system within 30 minutes of scram reset after the
first scram from full temperature
and pressure,
following each
refuelling outage.
In
a letter to
NRC dated
October
1,
1990, Subject:
BFN Safety Concerns
Associated with Pipe Breaks in the
System
(GSI 40 and Generic Letter 86-01), the licensee
stated that they revised
their Unit 2 abnormal
operating Instruction
(AOI) to require the
walkdown to be performed.
The licensee
also revised Unit 3 Abnormal
Operating Instruction 3-AOI-100-1, Reactor
Scram, to include this
requirement.
0
Il
il~
The inspector
reviewed Revision
5 of procedure
3-AOI-100-1, dated
July 10,
1995
and verified the requirement for walkdown inspection of
and instrument
volume to check for leakage
is specified
within the procedure.
The requirement
is contained
in procedure
Step
4.2. 15. 11, which states
that
a walkdown inspection of the
and
instrument
"should"
be preformed to check for leakage within 30 minutes
following scram reset,
of the first scram from rated temperature
and
pressure,
after
a refueling outage.
The inspector discussed
the need to
revise the procedure
by changing
"should" to "shall" so that it was
clear that the need to perform the walkdown of the
SDV is
a requirement,
and not optional.
The inspector
reviewed Unit 2 procedure
2-AOI-100-1, Reactor
Revision 40, dated
Hay 12,
1995.
Procedure
2-AOI-100-1, also states
that the walkdown of the
SDV "should" be inspected
following scram reset
after the first scram from full pressure
and operating
temperature
after
a refueling outage.
The inspector questioned
licensee
operations
personnel
regarding whether they have
implemented
the required
inspection after the first scram from full pressure
and temperature
during each of the three cycles since the
1991 restart of Unit 2.
The first scrams
from full pressure
and operating
temperature
following
the refueling outages
since the
1991 restart of Unit 2 occurred
on
August 2,
1991,
June
6,
1993,
and
December
2,
1994.
The information
provided
by the licensee,
(Unit 2 Reactor Operator Control
room logs for
these dates,)
was insufficient to determine if the licensee
had complied
with the commitment to perform the required
SDV system inspections
in
accordance
with intent of BWROG-08420.
Pending further review by NRC,
this issue
was identified as Unresolved
item 260/95-52-01,
SDV System
Inspection following Reactor
This is not
a Unit 3 restart
issue.
Within the areas
inspected,
violations or deviations
were not
identified.
3.0
3.1
Followup on Unit 3 Restart
Issues
Moderate
Energy Line Break
The moderate
energy line break
(HELB) evaluation
was performed to
determine
the effect of internal plant flooding outside containment
from
breaks
in moderate
energy lines (piping)
on safe
shutdown of the plant.
Moderate
energy lines are defined
as
systems
with pressures
less
than
275 psi or temperatures
less
than
200 degrees
F.
The Unit 3 evaluation
was performed in accordance
with the Unit 2 precedent.
The licensee
implemented
the Unit 3 HELB in a two phase
program.
Phase
I consisted
of a detailed
drawing review to identify all moderate
energy
lines which could be sources
of flooding in Unit 3, or common class
1
structures;
identification of flood compartments/areas;
identification
of potential
drainage
paths;
and identification of safe
shutdown
equipment
which could
be affected
by flooding.
The
Phase II evaluation
~
'i
included detailed calculations of flow rates
from various flood sources;
areas
affected
by flooding; maximum depth of water in the area;
and
drainage
paths
from the various flooded areas.
The inspector
reviewed the Bechtel
Report titled: Moderate
Energy Line
Break Flood Evaluation Report For Browns Ferry Unit 3, dated April,
1993.
This report contains
assumptions/conditions
for HELB analysis,
summary of design methodology,
references,
design input data,
design
analysis,
and
summary
and conclusions.
The report contained
a
discussion
on Browns Ferry's original methods for conformance
to AEC
requirements
for evaluating flood from HELB, and justification for not
including some
areas
in the
HELB analysis.
The conclusions of the
study were that Browns Ferry conforms to the original licensing basis
and that the existing flooding studies
and protective measures
are
adequate.
Considerations
in the
HELB analysis
included control of
flooding by providing drainage
paths
from areas
containing critical
equipment potentially suspectable
to flooding, use of curbs/barriers
to
protect vital equipment,
and mounting equipment
on pads
so that it
located
above potential flood elevation levels.
The inspector
concluded that the licensee's
MELB analysis
adequately
addresses
this issue.
The licensee's
design
assumptions
and design
methodology are technically adequate.
The
HELB program is acceptable
for restart of Unit 3.
3.2
Platform Thermal
Growth
The platform thermal
growth issue
involved the effect of thermal
loads
on structural
steel
platforms.
During review of structural
steel
design
criteria, the
NRC office of Nuclear Reactor Regulation
questioned
the
licensee
regarding their use of non-linear analysis
which predicted
plastic deformation of structures
due to thermal
loads.
As
a result of
these questions,
the licensee
performed
a comprehensive
review of their
design criteria
and concluded that they would revise the criteria to
require steel
members to remain within the elastic limit for all loading
combinations.
The licensee
submitted their revised criteria to
NRC for review.
A
Safety Evaluation Report
was issued
in
a letter to TVA dated
December
7,
1993, Subject:
Browns Ferry Nuclear Plant Supplement
Safety Evaluation
of Structural
Thermal
Growth Design Criteria
(TAC Nos.
M08618,
H80619,
and M80620), which accepted
the licensee's
long-term structural
steel
design criteria'
followup site audit was conducted
by
NRR on March
14
and
15,
1995, to review implementation of the licensee's
long term
design criteria.
During this audit, the
NRR staff examined
completed
modifications
on Unit 2 pl.atforms required
by the revised
thermal
load
design criteria.
The results of the audit are
summarized
in an
NRC
letter to TVA dated April 20,
1994, Subject:
Browns Ferry Nuclear Plant-
Audit of Structural
Steel
Design Criteria Implementation.
0
~i
The licensee
consi'dered
thermal
loads in their analysis of Unit 3
structural
steel
pl'atforms.
Hodifications required
by thermal
loads
included
used of slotted holes in beams,
addition of beam stiffeners,
cover plates,
etc.
These
were implemented
as part of design
change
packages
previously inspected
by NRC.
Issues
regarding modifications to Unit 3 structural
steel
platforms were
closed
by the inspector during the inspection
documented
in
NRC
'Inspection
Report
number 50-259,260,296/95-41.
This issue is resolved
for Unit 3 restart.
3.3
HVAC Duct Supports
'Numerous discrepancies, were identified in 1988 in the design
and seismic
qualification of heating, ventilation and air conditioning
(HVAC)
ductwork.
The licensee
developed
a program to inspect safety related
,HVAC duct work.
for Unit 2 restart,
the ductwork required for Unit 2
operation
was evaluated
to an interim operability criteria.
A safety
Evaluation Report
was issued
by
NRC on August 22,
1990 which concluded
that the Unit 2
HVAC ductwork and supports
were acceptable
for Unit 2
restart.
The Unit 2
HVAC ductwork and supports
were subsequently
qualified for long term requirements
of the .licensees
Design Criteria
BFN-50-C-7104.
For Unit 3,
a review was performed to identify class
I HVAC ductwork
which was not previously required for Unit 2 operation.
The only areas
identified as being specific to Unit 3, were the ductwork associated
.with the Unit 3 residual
heat
removal
(RHR)
and core spray
(CS)
pump
motor coolers.
This Unit 3 ductwork and supports
were evaluated
to the
long term requirements
of Design Criteria BFN-50-C-7104.
A design
change notice
(DCN) number
DCN W28617A,
was issued to implement
modifications to existing Unit 3
HVAC ductwork supports
and
added
one
new support.
The inspector
reviewed
DCN W28617A and performed
a walkdown inspection
to examine the
new support.
The inspector verified the
new support,
number 3-47B923-11,
was installed in accordance
with the design
requirements
shown
on drawing number 3-47B923-11,
Revision
1.
Acceptance criteria for installation of HVAC ductwork supports
are
specified in Hodification and Addition Instruction HAI-4.3,
HVAC Duct
'System,
Revision 9, dated
June
16,
1993.
The existing-support modifications consisted
of increases
in the size of
some fillet welds to
~k inch.
The inspector
reviewed quality control
inspection records
and weld data sheets for HVAC ductwork support
numbers
3-47B923-6,
3-47B923-7,
3-47B923-8,.
and 3-47B923-10
and verified
that the welds were modified as required
by the design drawings.
This
issue is resolved for Unit 3 restart.
0
3.4
Control
Rod Drive Hydraulic Piping System
During inspection of cable tray supports
in the Unit 2 reactor building
the licensee identified an issue regarding
attachment of control rod
drive
(CRD) system piping to the cable tray support structure.
The
licensee
performed
an extensive
design evaluation of the Unit 2
piping system
and
implemented modifications to the Unit 2
CRD pipe
support frames.
The licensee's
Unit 2
CRD pipe frame design reanalysis
program
was reviewed
by NRC during inspections
documented
in
NRC
Inspection
Report
numbers
50-260/89-20,89-31,
89-39,
89-44,
89-62,
90-08
and 92-01. This issue
was closed for Unit 2 restart
in NRC Inspection
Report 259,260,296/90-23.
A walkdown inspection of the Unit 3
CRD pipe support
frames
showed that
the Unit 3 frames,
were identical to the Unit 2
CRD frames.
Since the
Unit 2 frames required extensive modifications,
due to cost
and schedule
considerations,
the licensee
decided to replace
the Unit 3
CRD frames
with new supports.
The modification which involved installation of 32
new
CRD pipe support
frames
was
implemented
under
DCNs W17652,
W17653,
and W18645.
Three of the
new
CRD pip
support
frames
were inspected
during the inspection
documented
in
NRC Inspection
Report
number
259,260,296/95-03.
During the current inspection,
the inspector
inspected
an additional
five of the
new
CRD pipe support
frames.
The
new frames
were inspected
against
the design drawings for configuration,
member size,
weld size,
type and length connection details,
and other construction
requirements
stipulated
by the licensee's
procedures.
CRD pipe support
frames
inspected
were
as follows:
Support
numbers 3-47E-468-102,-103,-104,-
106,
and -107.
No discrepancies
were identified during the walkdown inspection.
The
inspectors
concluded that the modification .were implemented
in
accordance
with design requirements.
This issue is resolved for Unit 3
restart.
3,5
Large Bore Piping and Supports
The licensee initiated programs
in 1979 to comply with IE Bulletin 79-
02,
Pipe Support
Base Plate
Design Using Concrete
Expansion
Anchors,
and
Seismic Analysis for As-Built Safety-Related
Piping
Systems.
Implementation of these
programs
was delayed
by other
programs.
In addition,
the licensee did not include portions of piping
systems
covered
by other programs
under the
IEB 79-02/79-14
program.
Numerous deficiencies
were identified by
NRC in 1985
and
1986 concerning
implementation of these
programs.
In order to resolve
these
deficiencies,
the licensee
made various
commitments to
NRC regarding
improvement to design criteria, reinspection of large bore piping and
supports,
reanalysis
of piping and supports,
and implementation of any
required modifications.
Acceptance of the licensee's
design criteria
4I
>
for analysis of piping and pipe supports
by
NRC is documented
in NUREG-
1232,
Volume 3, Supplement
1, Safety Evaluation Report for Browns Ferry
Unit .2 Restart.
For Unit 3 restart,
the licensee
has
completed all 79-02
and 79-14 work
with the exception of a few pipe support modifications
on system
10, the
reactor
head vent.
These modifications will be completed after fuel
load for Unit 3, but prior to Unit 3 restart.
Inspection of the licensee's
79-02
and 79-14 program for 'Unit 3 included
review of pipe stress
analysis,
review of pipe support design
calculations,
and inspection of completed
pipe support modifications.
These
inspections
are documented
in NRC Inspection
Report
numbers
50-
259,260,296/91-34,
91-42,
92-07,
92-32,
92-38,
93-11,
93'-26,
93-29)
94-
15, 94-29,
and 95-03.
This issue is resolved for restart of Unit 3.
3.6
Long Term Torus Integrity
In the early 1980's the licensee
implemented
a series of modifications
to the torus intended to resolve deficiencies identified regarding
the
original design of the Hark
1 containment
system.
These modifications
involved torus attached
piping,
and structural
reinforcement of the.
torus
and torus related structures.
In 1985, discrepancies
were
identified by
NRC during inspections of the as-constructed
torus
attached
pipe support modifications.
The licensee's
corrective actions
included reinspection of the torus attached
piping and pipe supports,
and reinspection of torus structural modifications
and torus related
structures.
The licensee corrective action for Unit 2 restart
were
accepted
by
NRC in Section 2.2.4.4 of NUREG-1232,
volume 3, Supplement
2, dated
January
23,
1991.
In a letter to
NRC dated April 29,
1991, Subject:
Browns Ferry Nuclear
Plant
- Program for Resolving
Long-,Term Torus Integrity Issue Prior to
the Restart of Units
1
and 3, the licensee
provided
NRC their action
plan
and commitments for resolution of long term torus integrity for
Units
1 and 3.
The licensee's
corrective actions
included walkdown
inspections
to identify any discrepancies
in the torus, evaluation of
the discrepancies,
and,performance
of modifications to correct
any
unacceptable
discrepancies.
Inspect'ion of the licensee's
implementation of the long term torus
integrity program included review of design criteria, design
calculations
and completed modificati'ons for torus attached
piping and
pipe supports.
These inspection
are documented
in
NRC Inspection
Report
numbers
50-259,260,296/92-32,
94-15,
and 95-03.
The inspectors
concluded that the installed modifications were acceptable.
This issue
is resolved for restart of Unit 3.
il
0
3.7
3.7.1
- Cable Tray and Conduit Supports
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
cable tray and conduit supports,
and evaluation of existing cable tray
and conduit supports.
New Cable Tray and Conduit Supports
New supports
are those installed since
1986.
These
supports
are
designed
in accordance
with the licensee's
seismic design criteria and
installed
under the l.icensee's
quality assurance
program requirements.
The inspector
performed
a walkdown inspection
and examined
new cable
tray and conduit supports.
New supports
were inspected
against
the
design drawings for configuration,
member size,
weld size,
type
and
length,
connection details
and other constructions
requirements.
Additional acceptance
criteria utilized by the inspector during the
walkdown inspection
were Hodification and Addition Instruction, HAI-3.9,
Instal.lation,of Cable Tray Cable Tray Supports
and Cable Tray Covers,
Revision 7;
and HAI 3. 1, Installation of Conduit
and conduit supports,
Revision
25.
Cable tray supports
examined during the walkdown were:
Support
numbers
319092-15-B1102-38;
318992-1-81102-145;
318992-2-B830-56;
318992-3-
B1102-32;
318992-8-B1102-174;
3-48B1102-3-21,
-38, -122, -145,-164;
and
W17473-253,
-303,
and-304.
The fol.lowing deficiency was noted
by the inspector during the walkdown:
Two flare-bevel
on the vertical interfaces of Bill-of-Haterial
i,tems
3 and 4,
on cable tray support
number 319092-15-B1102-38,
had not
been
completed
as required
by the design details
shown
on drawing number
DCA W17473-300.
Paragraph
6. 1. 1 of Procedure
HAI-3,.9 requires
cable
tray supports
to be fabricated
and installed according to the applicable
design output documents
(drawings).
The failure of the licensee
to
install cable tray support
number in accordance
with the drawing
requirements
was identified as violati'on item 296/95-52-02,
Failure to
Construct
Cable Tray Support in accordance
with Design Requirements.
The licensee
issued
Problem Evaluation Report
(PER)
number
BFPER 951125
to document
and disposition this problem.
The inspector
also identified three cable tray supports
which had the
incorrect support
number
on the identification tag,
(Support
numbers
3-
-38,
and -145)
and
one support which had
a missing
identification tag (Support
number 318992-8-B830-15).
These
items were
also included
on
PER number
1
BFBER 951125.
However, since the mi'ssing
or incorrect tag numbers
have
no safety significance,
these
items were
not included
as. part of the violation.
~
~
During the walkdown, the inspector also
examined
a Kellems-grip cable
support,
support
number 3-4883800-4500,
which provides additional
'vertical support for cables
in vertical cable trays.
The inspector
noted
an 'error in the Bill of Materials
(BOH) for Items
1 through 5.
The licensee
issued
BF
PER 95-1128 to document
and disposition this
problem.
Further review of this problem disclosed that the error in the
BOM for items
1-5 was
due to
a drafting error in the drawing rollup
process.
This problem is similar to that identified in Unresolved
item
296/95-15-02.
Since the support
was constructed
as required
by design
requirements,
a violation was not identified for this problem;
however,
the drafting error was identified as
a weakness.
The following new conduit supports
were also
examined during the
walkdown inspection:
Typical conduit supports
number 0-48B805-010,
0-
and 0-48B805-014,
installed
on elevation
593; unique conduit
supports
number 3-48B3800-2188,
-4194,
and -4259;
and temperature
switch
support
number 3-47B900-212.
The conduit supports
were constructed
in
accordance
with design requirements;
however,
an additional drafting
error was noted
on drawing number 3-48B38004190,
Revision 0, regarding
incomplete/incorrect
drawing notes.
This problem was also documented
on
BFPER 951
128.
This was identified to the licensee
as another
example
of the weakness
discussed
above.
3.7.2 Evaluation of Existing Cable Tray and Conduit Supports
Seismic verification of existing cable tray and conduit supports
is
being accomplished
using the Generic
Implementing Procedure
(GIP) for
Seismic Verification 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.
The licensee
committed to
complete the A-46 walkdown for cable tray and conduit supports
in Unit 3
prior to restart of unit 3.
The inspector
reviewed walkdown Instruction
CEB-012,
Seismic Verification and Assessment
of Nuclear Plant
Equipment,
Revision 0, dated August 17,
1994.
This pro 'edure specifies
the
instructions for implementation of the GIP requirements
for personnel
qualifications, precautions,
methodology,
acceptance
criteria,
and
documentation
requirements.
The licensee
has
completed
the A-46 walkdown inspections for existing
cable tray and conduit supports
in all Unit 3 category I structures,
except for the drywell.
During the A-46 walkdowns,
the licensee
evaluate
cable tray fill, spans,
and supports,
including anchorage
and
conduit spans,
supports
and anchorage
using the criteria in GIP.
Cable
trays, conduits,
and supports
which did not meet the GIP acceptance
criteria were designated
as outliers.
The inspector
reviewed the results of the licensee's
A-46 walkdowns
summarized
in a walkdown summary table.
Outliers are documented
on
Outlier Seismic Verification Sheets.
The outliers are addressed
either
through
a plant work request,
or by a design evaluation
documented
in a
calculation using the GIP acceptance
and the licensee's
design criteria.
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Items addressed
by 'work requests
included missing or damaged
hardware
covered
by existing plant maintenance
procedures.
Problems (outliers)
which involved questionable
design
and/or construction practices,
e.g.
conduit over-spans,
apparent
inadequate
anchorages,
potential
seismic
interactions,
supports
which do not meet current design practices,
etc.
were evaluated
by the licensee
in calculation
number CD-(0000-931227,
Revision
1, dated
June 8,
1995, gualification of Cable Tray and Conduit
,System
by A-46 program.
Modifications
(DCNs) were issued for outl.ier
which could not be qualified.
The inspector reviewed the calculation for completeness,
accuracy
and
adherence
to design criteria and procedural
requirements.
No
deficiencies
were, identified.
The inspector walked down three,
randomly
selected,
modifications
implemented to resolve A-46 outliers,
and
verified the modifications were implemented
in accordance
with the
DCN
requirements..
Modifications examined
were
as follows:
- Modified Conduit support
on drawing number 3-48B3800-3935,
Revision
1
- Modified. Conduit Support
SL No.
3 on drawing number 3-48B3800-3936,
'Revision 0.
- New conduit support
as
shown
on drawing number 3-48B3800-3938,
Revision
0 ~
No deficiencies
were identified.
The inspector
conducted
a walkdown inspection
in the following Unit 3
areas
to assess
the effectiveness
of the licensee's
A-46 Cable Tray and
Conduit program.
Reactor Building Elevations
565,
593,
and 621; Control
Building cable spreading
room and elevation
593 hallway;
and Diesel
Generator Building.
No significant deficiencies
were identified,
however,
several
(18) minor items, were noted.
These
included missing or
loose lockouts,on cable tray hanger rods,
loose or mixed conduit clamps,
a temporary support still installed in the diesel
generator building,
broken conduits,
and other minor items.
The inspector also noted
housekeeping
deficiencies
such
as debris in cable trays,
missing cable
tray covers,
and .tools
and debris left in various
areas
in the plant.
Licensee
personnel
indicated that the housekeeping
deficiencies will be
addressed
as the work in the areas
is. completed
and the areas
turned
over to operations for restart.
In the areas
examined,
the inspector
concluded that the licensee's
A-46
cable tray and conduit walkdown program meets
NRC requirements
and is
adequate
for Unit 3 restart.
However, this issue will remain
open
pending completion of this program in the Unit 3 drywell.
In the areas
examined,
no deviations
were identified.
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10
4.0
Licensee
Event Report
(LER)
.(Closed)
Inadequate
Design Control
Procedures
,Discrepancies
in HVAC Ductwork.
In October,
1988
a review of open non-
conformance
reports identified several
discrepancies
involving the
design
and seismic qualification of HVAC ductwork.
This
LER was closed
for unit
2 restart
in
NRC Inspection
Report
number 50-259,260,296/91-06.
For Unit 3,
a review was performed to identify class
I HVAC ductwork
that
was not required for Unit 2 operation.
Modifications were
completed
as discussed
in paragraph
3.3,
above.
This
LER is closed for
unit 3 restart.
5.0
5.
1'.2
Licensee Action on Previous
Inspection Findings
(92701
and 92702)
(Closed) Violation Item 259,260,296,85-41-01,
Inadequate
Design Control
for Safety-Related
Cable Tray Supports.
This violation was identified in 1985
as
a result of a review of design
calculation for safety-related
cable tray supports
in various category
1
structures.
Problems identified included improper seismic design
analysis of various supports,
errors
and omissions
in the calculations,
and failure to perform design verifications.
This violation was issued
to the licensee
on September
8,
1986
as part of a
$ 150,000 civil penalty
covering
examples of failure to comply with NRC requirements
identified
in six
NRC inspections
covering the period from August
12,
1985 through
January
31,, 1986.
The licensee
did not contest
the civil penalty.
The licensee's
corrective actions for th'is violation are
stated
in
their letter to
NRC dated October 8,
1986, Subject:
and
Proposed
Imposition of Civil Penalty
Enforcement Action EA-86-56.
The licensee's
corrective actions
included preparation of procedures
and
design criteria for design of cable tray supports,
using
an independent
consultant to perform
an interim seismic qualification of Unit 2 cable
tray supports for Unit 2 restart,
and long-term qualification of cable
tray supports
using the Generic
Implementing
Procedure
(GIP) for Seismic
Verification of Nuclear Plant Equipment.
The GIP was issued
by the
Seismic gualification Utility Group in response
to
NRC Unresolved Safety
Issue
A-46 (USI A-46) Seismic Adequacy of Mechanical
and Electrical
Equipment.
The corrective actions for Unit 2 were reviewed during inspections
documented
in NRC Inspection
Report
numbers
50-259,260,296/88-38,
89-21,
89-29,
89-30, 89-32, 89-42,
and 89-62.
The inspections
covered review
of design criteria, design calculations,
and selected
cable tray and
conduit support modifications.
(Closed)
Inspector
Followup Item 259,260,296,85-51,-01,
Inspection of
Existing Cable Tray Support
Systems.
This IFI was identified in 1985 during
a followup inspection
performed
relative to violation item 259,260,296/85-41-01.
The inspector noted
~
'l
ik~
11
5.3
that the licensee
did not have
a written procedure
to inspect existing
cable tray support
systems
to assure
the as-built cable tray support
systems
comply with applicable
design
documents.
The licensee
is in the
process
of inspecting existing cable tray support
systems,
and other
mechanical
and electrical
equipment,
using the
NRC approved
Generic
Implementation
Procedure
(GIP) for Seismic Verification of Nuclear Plant
Equipment,. referenced
above.
The licensee
issued
walkdown Instruction
CEB-012,
Seismic Verification and Assessment
of Nuclear Plant Equipment,
to implement the walkdown,program using the GIP.
The GIP contains
specific requirements
pertaining to inspection,
evaluation
and
.identification of cable tray support
systems
w'hich do,not meet seismic
design requirements.
(Closed)
Unresolved
Item 296/86-06-02,
Reactor Building Control
Bay
Inadequate
Design.
This item concern
the licensee's
identification of inadequate
design of
HVAC support.
'This issue
was reported to
NRC under
This unresolved
item was closed for Unit 2 in NRC Inspection
Report
number 50-259,260,296/90-08.
For Unit 3,
a review was performed to
identify the class
I HVAC ductwork that was not required for Unit 2
operation.
The only areas specific to Unit 3 were the duct work for the
Unit 3
RHR and core spray
pump motor coolers.
The licensee's
corrective
action for this duct work is discussed
in paragraph
3.3,
above.
296/86-06-02's
closed for Unit 3.
URI 259/86-06-02 will remain
open
for Unit 1.
5.4
(Closed)
Unresolved
item 296/87-26-03,
RHR Pump Suction
and Nozzle
Load
Allowable Are Exceeded.
This item concerned
Residual
Heat
Removal
(RHR) Nozzle load allowables
as identified by the licensee
in deficiency number 87-13-6 of
Engineering
Assurance
Audit 87-13.
The licensee
revised calculation
number 'CD-(3073-920014
(System Nl-373-5R)
and generated
new calculation
number CD-f3074-910631
(System Nl-374-5R)
and CD-g-3074-910400
(System
Nl-374-7R) to evaluate
the
RHR pump suction anchor
and nozzle loads.
The revised
and
new calculations qualified the applied loads
based
on
revised design criteria BFN-50-C-7103,
General
Design Criteria for
Structural
Analysis
and gualification of Mechanical
and Electrical
Systems
(Piping
and Instrument Tubing).
The applied loads
include I.E.
Bulletin 79-14 requirements.
The as-built walkdown information and data
also were used in the analysis.
The inspectors
reviewed the following calculations
which qual.ified the
nozzle loads:
- Calculation
number CD-93074-910631,
Revision 4, dated
February
27,
1995.
- Calculation
number CD-(3073-920014,
Revision
11, dated .February
2,
1995.
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Oi
12
- Calculation
number CD-(3074-9104000,
Revision
5, dated April 5,
1995.
Based .on this review, the inspectors
determined that the calculations
complied with the licensee's
design criteria and were acceptable.
The
piping stresses
are within code allowable values.
This item is closed
for Unit 3.
URI 259/87-26-03 will remain
open for Unit l.
6.0
Exit Interview
The inspections
scope
and results
were summarized
on August
25 and
September
14,
1995, with those
persons
indicated in paragraph
1.
The
inspectors
described
the areas
inspected
and discussed
in detail the
inspection results listed below.
Proprietary information is not
contained
in this report.
Dissenting
comments
were not received
from
the licensee.
Unresolved
Item 260/95-52-01,
SDV System
Inspection
Following Reactor
paragraph
2.0.
Violation Item 296/95-52-02,
Failure to Construct
Cable Tray Support in
Accordance Mith Design Requirements,
paragraph
3.7. 1.
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