ML18033B270
| ML18033B270 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/17/1990 |
| From: | Carpenter D, Little W, Patterson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18033B268 | List: |
| References | |
| 50-259-90-05, 50-259-90-5, 50-260-90-05, 50-260-90-5, 50-296-90-05, 50-296-90-5, NUDOCS 9004250256 | |
| Download: ML18033B270 (35) | |
See also: IR 05000259/1990005
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
RePort
NoseI
50-259/90-05,
50-260/90-05,
and 50-296/90-05
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 Units 1, 2, and
3
Inspection at Browns Ferry Site near Decatur,
Inspection
Condu ted:
February
16 - March 16,
1990
~+~-c-'. +8~ c~g
Q~ Dr
arpe ter
NRC Si
e Manager
P//~~a
g
p,<
C.
.
atterson,
NR
estart
oor inator
Accompanied
by:
E. Christnot,
Resident
Inspect or
W. Bearden,
Resident
Inspector
K. Ivey, Resident
Inspector
R. Bernhard
Project Engineer
Approved by:
W. S.
ttle, Section
C ief,
Inspe tion Programs,
TYA Projects Division
Da e
gned
ate
igned
SUMMARY
Scope:
This routine resident
inspection
included surveillance observation,
maintenance
observation,
review of ECP subcategory
report operational
safety verification,
modifications, restart test
program,
action
on previous
inspection findings,
and licensing activities.
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Results:
A violation was identified for failure to maintain operable fire hose stations
as required
by TS.
This was in addition to two previous violations
and several
LERs
during
the
past
two years
concerning
compensatory
fire protection
measures.
A system
and
procedures
were in place
to correctly identify and
control
compensatory fire protection
measures,
but management
control of the
system
was not effective,
paragraph
5.a.
A violation for failure to verify power available during
an electrical
power
source
transfer
was identified,
paragraph
5.c.
Although the defueling of
Unit 2
was carried out in
a controlled
and methodical
manner,
the licensee's
review of the
cause
of a bent refueling
boom revealed
procedural
and training
inadequacies,
paragraph
5.b.
Ouring
1989
and continuing in 1990 replies
were not
timely and
frequent
extensions
were
requested
to the
time requirements
of
paragraph
9.
0
REPORT
DETAILS
Persons
Contacted
Licensee
Employees:
0. Zeringue, Site Director
- G. Campbell, Plant Manager
M. Herrell, Plant Operations
Manager
R. Smith, Project
Engineer.
J. Hutton, Operations
Superintendent
A. Sorrell, Maintenance
Superintendent
G. Turner, Site guality Assurance
Manager
P. Carier, Site Licensing Manager
- P. Salas,
Compliance Supervisor
J. Corey, Site Radiological Control Superintendent
R. Tuttle, Site Security Manager
Other
licensee
employees
or contractors
contacted
included
licensed
reactor operators,
auxiliary operators,
craftsmen,
technicians,
and public
safety officers; and quality assurance,
design,
and engineering
personnel.
NRC Employees
- W. Little, Section Chief
- C. Patterson,
Restart Coordinator
- E. Christnot,
Resident
Inspector
- W. Bearden,
Resident
Inspector
- K. Ivey, Resident
Inspector
- Attended exit interview
Acronyms used throughout this report are listed in the last paragraph.
Surveillance Observation
(61726)
The inspectors
observed
and/or reviewed the SI procedures
discussed
below.
The inspections
consisted
of a review of the SIs for technical
adequacy
and
conformance
to
TS, verification of test
instrument calibration,
observation
of the conduct of the test,
confirmation of proper
removal
from service
and return to service of the system,
and
a review of the test
data.
The inspector
also verified that limiting conditions for operation
wer e met, testing
was
accomplished
by qualified personnel,
and the SIs
were completed at the required frequency.
The
inspector
observed
portions
of O-SI-4.9.Al.a(A),
"Monthly
Operability on
1A Diesel Generator".
No deficiencies
were identified
with the performance of this SI.
0,
The
inspectors
observed
portions
of the
preparations,
equipment
setup,
and performance of 3-SI-4.9.Al.d(3A), "Annual Inspection
on
3A
Diesel
Generator".
This surveillance
consists
of making routine
internal
components
inspections
for the diesel
generator
and took
approximately five days
to perform.
The only problem that the
licensee
experienced
was
a
PMT deficiency identified when the diesel
failed to obtain the required
speed of 900 - 910
RPM.
The indicated
speed
was
897
RPM and this
was evaluated
and found to be
a problem
with the tachometer calibration.
The tachometer
was recalibrated
and
the
PMT considered
acceptable.
No violations or deviations
were identified in the Surveillance
Observation
area.
3.
Maintenance
Observation
(62703)
Plant
maintenance
activities
on
selected
safety-related
systems
and
components
were observed/reviewed
to ascertain
that they were conducted
in
accordance
with requirements.
The following items were considered
during
this review:
limiting conditions for operations
were met; activities were
accomplished
using
approved
procedures;
functional
testing
and/or
calibrations
were
performed prior to returning
components
or systems
to
service;
quality control
records
were
maintained;
activities
were
accomplished
by qualified
personnel;
parts
and
materials
used
were
properly certified;
clearance
procedures
were
adhered
to;
Technical
Specifications
were
met;
and radiological controls
were
implemented
as
required.
Maintenance
requests
were reviewed to determine
the status
of
outstanding
work activities
and to assure
that priority was assigned
to
safety-related
equipment maintenance
which might affect plant safety.
An inspector
observed
portions of the annual
inspection for DG 3D
(PM F1672).
This work was
performed in accordance
with MMI-6, "Scheduled
Maintenance of the Standby
Emergency
Diesel
Generator
Engines," Section
C.
This
was
part of the
planned
PM program for the
DG engines.
No
deficiencies
were identified.
An inspector
observed
portions of the replacement
of the outlet scram
valve actuator
on the Unit 2 "east side" hydraulic control
units.
This
work was
performed
in accordance
with MCI-0-085-VLV003,
"Outlet
Valve Disassembly,
Replacement,
Actuator
Replacement,
Valve Seat
Replacement,
and Valve Reassembly".
No
deficiencies
were identified.
An inspector
observed activities associated
with electrical signal tracing
and
reviewed
associated
(see
paragraph
6.c).
No deficiencies
were
identified.
No violations or deviations
were identified in the Maintenance
Observation
area.
Review of ECP Subcategory
Report
An inspector
reviewed
ECTG subcategory
report 31000, "Operations/
Operational," to determine if any of the employee
concerns
were applicable
to Browns Ferry
and what corrective actions
were taken.
Element
31003
included
an
employee
concern,
IN-86-055-003,
involving the
improper
control of root valves for tygon tubing used for temporary level
indication at Watts
Bar.
The subcategory
report determined
that this
employee
concern
was applicable
to Browns Ferry because
tygon tubing is
used for temporary vessel
level monitoring during
a vessel
drain down for
jet
pump work and recirculation riser piping repair work.
CATD 310.03-
BFN-01
was
issued
to revise
Standard
Practice
BF 14.25,
"Clearance
Procedure",
to require
that
tubing
used
for temporary
level
indication
be controlled
by
a caution order.
This item was closed
by ECP
reviewers.
The procedure
revision
was verified by
NRC inspection
(IR 88-22); however,
the inspectors
determined that the procedure revision
did not address
control of the root valves,
which was the initial problem.
During this inspection,
the inspector
noted that the current
equipment
clearance
procedure,
SDSP
14.9,
did not contain
the requirement for a
caution
on tygon tubing, nor did it address
the control of root valves for
tygon tubing.
From a review of an updated
CATD and associated
documentation,
and
discussions
with licensee
personnel,
the
inspector
determined that the licensee
had deleted
the requirement during
a general
revision of the procedure.
Furthermore,
licensee
personnel
stated that
the caution
was
not necessary
since
the
program which is in place
to
control
temporary
changes will provide
more appropriate
control
than
a
caution order.
PMI-8.1, "Temporary Alterations", provides administrative
control for all temporary alterations
at
Browns Ferry.
This procedure
includes
reviews
of
a
TACF by Technical
Support
and Operations,
and
approval of the
TACF by the
PORC.
The inspector
concluded that the
TACF
process
should adequately
ensure
the control of temporary tygon tubing at
Browns Ferry.
The
ECTG utilized
CATO's
to
ensure
that
corrective
actions
were
implemented for valid employee
concerns.
Even
though the actions
were
acceptable
for this
employee
concern,
removing
the requirement
for
a
caution order from the clearance
procedure after closure of CATD 310.03-
BFN-Ol is of concern
to the
NRC.
It appeared
that the revision was
made
without addressing
the fact that the requirement
was
a
ECP corrective
action.
Licensee
personnel
indicated that
SDSP
2. 1, "Site Procedures",
was revised
in June
1988 to require that
a note is included with each
procedure
step
associated
with
ECP corrective action commitments.
This
revision should provide assurance
that procedural
corrective actions
taken
since
June
1988 are maintained;
however, this is not true for procedural
corrective actions
implemented prior to June
1988.
The inspector brought
this issue to the attention of licensee
management
at the exit meeting
on
March
16,
1990.
The licensee
stated
that they would develop
an action
plan
and work with the
NRC Resident
Inspectors
to resolve this concern.
This issue
is identified
as
IFI 259,
260,
296/90-05-01,
ECP Corrective
Actions.
No violations or deviations
were identified during
the
review of
Subcategory
Reports.
5.
Operational
Safety Verification (71707)
The
NRC inspectors
were kept informed of the overall plant status
and any
significant safety matters related to plant operations.
Daily discussions
were held with plant management
and various
members of the plant operating
staff.
The
inspectors
made routine visits to the control
rooms.
Inspection
observations
included
instrument
readings,
setpoints
and
recordings;
status
of operating
systems;
status
and alignments of emergency
standby
systems;
onsite
and offsite
emergency
power
sources
available
for
automatic
operation;
purpose of temporary
tags
on equipment controls
and
switches;
alarm status;
adherence
to procedures;
adherence
to
limiting conditions
for operations;
nuclear
instruments
operability;
temporary alterations
in effect; daily journals
and logs; stack monitor
recorder traces;
and control
room manning.
This inspection activity also
included numerous
informal discussions
with operators
and supervisors.
General
plant tours
were conducted.
Portions of the turbine buildings,
each reactor building, and general
plant areas
were visited.
Observations
included
valve
positions
and
system
alignment;
and
hanger
conditions;
containment
isolation
alignments;
instrument
readings;
housekeeping;
proper
power supply
and breaker
alignments;
radiation
area
controls;
tag controls
on equipment;
work activities
in progress;
and
radiation
protection
controls.
Informal
discussions
were
held with
selected
plant personnel
in their functional areas
during these tours.
a
~
Fire Protection
Problems
(1) Observation
During
a routine tour of the reactor building on February 27,
1990,
the
inspector
observed
two
signs
that
appeared
to
contradict
each other at
a fire h'ose station.
The station
was
in the Unit 2 reactor building at fire hose
R2-6.
One sign
stated
"HOSE
STATION
OUT
OF SERVICE,"
and
the other
stated
"COMPENSATORY
FIRE
HOSE
FOR
UNIT
1
ELEVATION 639."
The
inspector
observed
a gated
wye connection in Unit 3 connected
to
a single
50 foot roll of hose,
and that the fire hose station in
Unit
1 was'lso
out of service.
Similar conditions
were
on
elevation levels
621,
593,
and 565.
The'nspector
went to the
control
room
and
discussed
the conflicting signs
with the
operations staff.
The operator
was unable to explain the signs
and initiated
a phone call for an explanation.
(2)
(3)
(4)
Removal
from Service Permits
The inspector
asked
to review the fire protection
removal
from
service permits,
but they were not in the control
room.
These
permits
are
termed
"Attachment
F"
and list the
compensatory
requirements
when fire protection
equipment
is
removed
from
service.
The Fire Protection
Equipment
and Barrier Penetration
Removal
from Service
Permit
(Attachment
F) are
processed
in
accordance
with procedures
FPP-2, "Fire Protection-
Attachments".
The book containing the Attachment
Fs
was located
in the operations
work control center
two elevations
below the
control
room.
The inspector
reviewed the
book but only a four
page
summary of the Attachment
Fs
was in the
book, without
a
description of the compensatory
measures.
The inspector
went to
the fire protection building and found the Attachment
Fs.
The
signs represented
a condition in the reactor building where
the
hose station for Unit 3 was
being
used for Unit 2 and the
hose for Unit 2 was being used for Unit 1.
Upon
a fire in
Unit 1,
a
hose
from the gated
wye connection
in Unit 3 would be
extended
to the Unit 2 hose station,
the hose at Unit 2 would
have to be disconnected
from the station
and connected,
and then
the Unit
2
hose
extended
to Unit 1.
The inspector
discussed
with
a
member of operations
management
on February
28,
1990,
that operations
personnel
seemed
unaware
of the
compensatory
fire measures
and the
TS responsibility was in effect delegated
to the fire protection staff.
TS Requirements
The inspector
reviewed the
TS requirements
of 3. 11.E concerning
fire hose
stations.
TS 3. 11.E requires
that the fire hose
stations
shown
in Table
3. ll.c shall
be
whenever
equipment
in the
areas
protected
by the fire hose stations
is
required to be operable.
When
a hose station is inoperable,
a
gated
wye shall
be
connected
to the
nearest
hose
station.
One outlet of the
wye shall
be
connected
to the
standard
length of hose
provided for the
hose station.
The
second outlet of the
wye shall
be connected
to a length of hose
sufficient to provide coverage for the area left unprotected
by
the inoperable
hose station.
Also, where it can
be demonstrated
that the physical
routing of the fire hose would be
a hazard,
the fire hose shall
be stored
in
a roll at the outlet of the
hose station.
Signs shall
be mounted
above
the gated
wyes to identify the proper hose to use.
Violations
On March 1,
1990,
the inspector
discussed
with the Operations
Manager that the plant was outside
It was discussed
that the hoses
were not connected
or routed,
and the adequacy of
using
a single Unit 3 hose station for compensatory
measures
for
both Unit
2 and Unit 1.
This
was
an example of violation of
TS 3. 11.E in that the inoperable fire hose stations
were not
connected
to an outlet of a gated
wye connection with sufficient
length
of
hose
to
provide
coverage
for the
areas
left
unprotected
by the inoperable
hose stations.
This is the first
example
of violation
259,
260,
296/90-05-02,
Inadequate
Compensatory
Fire Protection
Measures.
Action was initiated by operations
management
to review the TS.
The licensee
found that the existing hold order (0-90-160) for
Units
1
and
2 had
been
extended
to include Unit 3 on March 1,.
1990
due to substantial
leakage
through
the isolation valves.
There
were
no operable
hose stations
in the Units 1, 2,
and
3
reactor building.
This is the second
example of violation 259,
260, 296/90-05-02.
Licensee Corrective Actions
The licensee initiated actions
to restore
the Unit 3 system to
operation.
This
was completed
on March 5,
1990.
This issue
was
discussed
with all the shift supervisors
on March 2, 1990.
Hot
work activities
(welding,
grinding,
etc.)
in the
reactor
building were
stopped.
The fire truck
and fire protection
personnel
were
placed
on standby.
Sufficient hose
length
was
connected
to Unit 3 hose stations
to reach
the other units.
The
licensee
stated
the
hoses
were not routed
due to the
number of
modifications activities occurring in the plant.
Safety Assessment
During the review of this event it was determined that the fire
suppression
systems
were
for Units
3
and
1
and
for Unit 2.
A source
of water
was
available,
although
TS 3. 11.E requirements
were not met.
All three units
were
defueled
at
the
time,
which minimizes
the
equipment
required to be in service
and
reduces
the portions of the fire
protection
system required to meet TS.
The
above
violation
(90-05-02)
when
considered
with other
violations identified
by the
NRC
and
TVA over the
past
two
years
raises
concern
about
management
attention
to the fire
protection
program at Browns Ferry.
The other problems
are
as
follows:
Violation 89-33-04,
Breach of Fire Door.
This involved one
example
of failure to
implement
compensatory
measures
required
by
TS
LCO 3. 11.G.1.a for inoperable fire rated
assemblies.
Violation 89-43-04,
Failure to Maintain
TS
Compensatory
Neasures
for Inoperable
Fire
Hose Stations.
This involved
one
example of failure to maintain
compensatory
measures
required
by
TS
LCO 3. 11.E.l.a for inoperable fire hose
stations.
This
apparent
violation
was
subsequently
included
in
a Severity
Level III violation of the
program.
Violation of Fire Protection
Technical
Specification
Due to Personnel
Error.
This involved four
examples
of failure'o
implement
compensatory
measures
required
by
TS
LCO 3.11.G for blocked
open fire doors
without operable fire detection
systems
on either side of
the doors.
Plant Technical
Specification
Surveillance
Requirement
Exceeded
Due
to
a
Nisinterpretation
by
Supervision
Responsible
for Patrolling Firewatches.
This
involved
one
example of continuous
failure to implement
compensatory
measures
required
by
TS
LCO 3. 11.A. l.b for
fire protected
zones or areas with inoperable detectors.
The inspector
concluded that
a system
and applicable
procedures
were in place
to correctly identify and control
compensatory
fire protection
measures.
However,
management
control of the
system
was
not effective
and there
appeared
to
be
a lack of
adequate
knowledge of the compensatory fire protection in place
at any given time by the plant operations staff.
The licensee
is requested
to address
these
concerns
in their response
to
Violation 90-05-02.
The focus of the review was
the reactor building and did not
include
a detailed
assessment
of other plant areas
and other
times the violation may have occurred.
(7)
Other Locations
At the
same
time
the
Units
1
and
2, reactor
building
hose
stations
were
removed
from service,
the
common Units
1
and
2
Diesel
Generator
Building fire protection
was
removed
from
service.
Fire hoses
were routed from a fire hydrant.
There is
no provision for this in TS 3. 11.E, although this appeared
to be
a reasonable
method of providing fire protection
to the
D/G
Building.
Nevertheless,
as
a conservative
measure,
hoses
were
connected
to an operable
Unit 3 hose station in elevation
1C of
the control
bay for routing
down the corridor and into the
D/G
building.
(8)
Diesel Drive Fire
Pumps
Tour
On
Narch
14,
1990,
three
NRC inspectors
accompanied
by fire
protection
and operations
personnel
toured
the diesel
driven
fire
pumps facilities
on site.
The fire protection
system
e
consists
of an inside
loop consisting of three electric fire
pumps
and
a
2500
gpm diesel
driven fire pump supplied
by river
water.
An outside
loop consists of a 300,000 gallon bladder
and
associated
2000
gpm diesel
driven fire pump located outside the
and
a 250,000 gallon bladder
and associated
1500
gpm
diesel
driven fire
pump
located
at
the
low level
radwaste
storage facility.
The outside
loop is supplied
by potable water
from the City of Athens.
The licensee
also
has
two fire trucks
of 1250
and
750
gpm capacity.
A connection
existed
on the
outside of the
2000
gpm diesel
driven fire pump building for
connecting
a fire truck to
the
300,000
gallon
bladder.
Connections
at the outside of the
2000
gpm diesel
drive fire
pump building
and
a fire hydrant
near
the east
gate
portal
provided
a means
to connect
hoses
to tie the outside loop to the
inside loop as
a backup supply of water.
The inspector
noted during the tour that the
2500
gpm diesel
driven fire pump building had signs stating that smoking
was not
allowed within 50 feet.
Signs were not observed
on the outside
of the building at
1500
gpm
pump.
No other
concerns
were
identified.
The inspector
concluded that the facility has
the
capability to supply
a backup
supply of water if required
by TS 3.11.B.
b.
Final Event Report
on Bent Unit 2 Refueling Platform
Boom
The inspector
reviewed the final event report 90-002,
concerning
the Unit 2 refueling platform boom.
This event occurred while
defueling Unit 2 and delayed
the offload for approximately three
days.
On January
9, 1990, at 4: 10 p.m., while moving the refuel
bridge across
the
SFSP,
the fuel grapple
came in contact with a
blade guide handle
and the
boom was bent.
The root cause of this event
was that
no procedural
guidance or
training existed
which specified
minimum grapple
height for
traversing
the
SFSP with an unloaded
grapple or for moving the
grapple
in
more
than
one
dimension
at
a time.
Possible
contributing factors were
an undocumented
modification which may
have resulted in unreliable depth indications,
and
a blade guide
which may not have
been fully seated
in its storage
location.
The inspector
noted that the defueling activities were carried
out in a controlled
and methodical
manner.
The final report was
thorough
and critical to point out the procedural
and training
inadequacies.
0
Failure to Follow Procedure
On March 1,
1990,
the Unit 3
Bus
3A was deenergized
during
manual
transfer
of
480
Volt Shutdown
Board
3A from the
associated
alternate
power source
to the
normal
power source.
This resulted
in unplanned
automatic initiation of all three
Standby
Gas
Treatment
System
trains
and
both
Control
Room
Emergency
Ventilation trains.
Additionally, isolations
were
received
on the Unit 3 Reactor
Zone Ventilation System;
Refuel
Zone Ventilation Systems
for all three units;
and Unit 3
Groups 2,3,6
and 8.
480 volt Shutdown
Board
3A had
been
aligned to the alternate
power
source
due to planned
preventive
maintenance
activities
associated
with the
3A DG and the 4160 volt and 480 volt circuit
breakers
that provide normal electrical
power to that bus.
When
the operator
attempted
to transfer
the power source
back to the
normal supply power, the shutdown
board
was lost due to the
4
KV
feeder
breaker
to the normal
supply breaker
being open.
During
the licensee's
post event evaluation it was determined that the
licensee's
configuration control
records
did not reflect the
fact that the
breaker
was
out of position,
a condition that
could
have
been
a contributing factor to the event.
However,
O-OI-57B, 480V/240V AC Electrical
System Operating Instructions,
Step 8.6.3, clearly requires that the normal feeder
breaker
Volts indicate greater
than
450 volts prior to transferring the
board
power supply.
The transfer activities were being directed
by
a Senior Reactor Operator licensed
ASOS
and performed
by an
electrically qualified assistant
unit operator.
Additionally,
the
voltage indication for the electrical
power
sources
are
located
on 480 Volt Shutdown
Board
3A and are easily within view
of the individuals making the transfer.
This failure to follow procedure
does
not meet the requirements
to qualify as
a
NCV due to poor performance
in the
area of
failure to follow procedures.
Therefore,
this violation will
not
be considered
a
NCV.
Similar events
described
in Unit 2
LERs 88-04, 88-07,
and 89-05 are associated
with personnel
error
or failure to follow procedures
that resulted
in unplanned
actuations
or scrams.
Unit 2 LER 88-08 was very similar in that
an unplanned
ESF actuation
occurred
when electrical
power to the
2B 480 volt Shutdown
Board
was lost during transfer
due to an
-open
4160 volt normal
supply breaker
being
open.
During the
inspectors
review of licensee corrective actions associated
with
this
item as
documented
in
IR 89-27, it was
noted that the
licensee
had taken action to prevent recurrence
including:
Individual counseling of operations
personnel
involved in
the event focusing
on the necessity for strict attention to
detail
when performing assigned
tasks.
e
10
Since
a contributing factor in this
event
was
a test
procedure
which
did
not
include
adequate
specific
instructions,
members
and alternates
were
provided
guidance
to
ensure
that
procedures
included
adequate
direction for returning
systems
to normal
upon completion
of the test.
This failure
was
identified
as
a violation of
TS 6.8. 1.1.a
(Violation
259,
260,
296/90-05-03,
Failure
to Follow Operating
Instruction).
d.
Unit Status
All three units remain defueled
and in an extended
outage
as part of
the
BFNP recovery plans.
Work activities for returning Unit 2 to
service substantially
increased
near the
end of this report period.
The main activities were completion of pipe support
and restraints.
6.
Modifications (50090,
51063)
a ~
Workplan Review
The inspector
reviewed
the following completed
workplans which were
stored in plant lifetime storage.
WP
2600-89
involved the replacement
of relays
in
RPS Circuit
Protector
Cabinets
2C1
and
2C2.
This
work
was classified
safety-related
but not affecting
EQ components.
No problems
were identified with the workplan.
WP 2588-89
involved the installation of
EQ terminal
blocks in
MOVs 2-FCV-74-106
and
2-FCV-75-39
and the deletion of unused
terminal blocks in MOVs 2-FCV-74-71
and 2-FCV-75-30.
During the review of WP 2588-89 the inspector
noted that the workplan
included the requirement
to use only Marathon
300 terminal blocks or
a qualified splice to replace existing terminal blocks.
The work was
properly
documented
on
a
Form
235 in accordance
with
SDSP
7.7,'ualification
Maintenance
Data Sheets
(QMDS) Implementation
and Harsh
Equipment Maintenance
System.
e
The inspector also noted that revision
1 to this
WP added additional
steps
31.0
through
34.0 which removed/replaced
jumpers
associated
with 2-FCV-75-39 which were too short or cut.
This replacement
was
performed in accordance
with MAI - 3.3, Attachment 5, Cable/Wire Lift
Data Sheet.
The inspector
reviewed
MAI 3.3,
Cable Termination
and
Splice for Cables
Rated
up to 15,000 Volts,
Rev.
5, which was in
effect at the time of the above work and determined that
Attachment 7, Internal
Panel
Wire/Jumper
Data Sheet,
should
have
been
used rather
than Attachment
5 for this work.
Attachment
7 included
the provision for recording
new wiring'ype, Mark Number,
and reel
number
and
the
performance
of point to point continuity checks.
These
were not included
on Attachment 5.
The inspector
held discussions
with members of the licensee
Quality
Organization
to discuss
the
above deficiency.
After performing
a
separate
review of the subject
workplan the quality organization
representative
acknowledged
that the incorrect attachment
was
used
during performance of the workplan.
However, during the licensee's
subsequent
review of the workplan,
the licensee
Quality Evaluator
identified that the missing
items not documented
by an Attachment
7
were actually
performed
by separate
activities
documented
in the
workplan.
The inspector
noted that the licensee
performed
EMI - 18,
Limitorque Switch Adjustment for High Speed
CSSC
8
Non-CSSC Valves,
as post-maintenance
testing.
As the result of the inspectors
concerns
in this area. the licensee's
Quality Organization initiated
a special
Quality Monitoring Report in
this
area
which selected
for review
a
minimum of six to eight
workplans that were performed in the
same period to determine if this
was
a generic
problem.
Each of the associated
workplans
used
the
correct attachment.
The inspector
concurred with the licensee's
position that the
above
inadequate
work instruction constituted
an isolated
case
and
a minor
deficiency with no safety significance.
b.
Field Activities - Pipe Supports
and Restraints
Systems
1)
Pipe Supports
EECW System
The
inspector
observed
the activities
associated
with pipe
supports
for the
EECW,
and
Core
Spray
Systems.
The
inspector
noted that for the
EECW system approximately
40 DCNs,
83
WPs
and
750
supports
were
involved with the
overall
modifications effort inthis area.
The total
amount of time
dedicated
by the licensee for this work was approximately three
months.
The following specific activities were reviewed:
The
EECW pipe support activities in the Unit
1 area
involving the north header of EECW.
The
EECW pipe support activities in the Unit 2 area
involving the north and south
of EECW.
The
EECW pipe support activities in the Unit 3 area
involving the south
of EECW.
The
inspector
noted
that all activities
observed
were
in
accordance
with applicable
WPs
and
QC inspector activities were
present.
0
12
An inspection
was performed of the licensee's
Warehouse
No.
14,
where pipe support material
was staged.
A significant amount of
rust
was observed
on
some of the support material
dedicated
to
specific support activities in the field.
This observation
was
discussed
with senior
TVA management
and
TVA gA supervisors.
2)
Pipe Supports -
RHR System
The inspector
reviewed
ECN/DCNs involving pipe supports
on the
RHR systems.
A total of 34
ECN/DCNs were identified as
being
field complete,
in progress,
or in review status.
The inspector
reviewed
DCN
numbers
W8850A
and
W9364A.
The following was
observed:
DCN W8850A
Involved the installation,
removal or
modification of nine
pipe supports
located
in
Unit
1 on
RHR Loop II.
DCN W9364A
This modification involved seismically qualifying
the
RHR system of Unit
1 Loop II.
In order to
limit the
amount of seismically qualified piping
and associated
supports,
this modification
adds
blind flanges
to the
RHR system cross-tie piping
at column R4.
This
DCN provides for the
installation of blind flanges,
and
a vent line
with associated
isolation valves
on the Unit
1
RHR system cross-tie line.
This is
a
modification to the Unit
1
RHR system to support
Unit 2 operation.
The
inspector
walked
down the
areas
where
the
pipe support
activities are scheduled
to occur in the Unit
1
RHR room.. The
inspector
noted
no deficiencies
in the areas
reviewed.
Electrical Signal
Cable Tracing
The licensee
conducted
electrical
signal
tracing
on
13 cables
as
requested
by the Electrical
Issues
Inspection
Team and documented
in
IR 89-59.
The methodology
used
by the licensee
involved the issuance
of
13
MRs,
numbers
1024810
through
1024822
which designated
the
cables
to
be traced.
The inspector
reviewed the
MRs as
issued
and
observed activities in, the field.
Although all
MRs had not been closed out at the end of this reporting
period, all activities
observed
appeared
to be in keeping with the
inspection
team's
request.
The specific field observation
consisted
of observing
the licensee
perform
cable
routing using
procedure
SEMI-62, Revision
2,
Cable
Route Verification.
This activity consisted of using
a
RD 600 Radio
0
13
Detection device,
set to beep
on
8
kHz signal
hooked
up to
a cable
conductor
and tracing
the
path of the
cable
using
a
hand
held
receiver.
d.
ECN/DCN Review
The inspector
reviewed
ECNs/DCNs for the following programs.
I)
Appendix
R
The
inspector
reviewed four
ECNs/DCNs
P0882,
0879,
0819
and
5289.
These
resulted
in the initiation of the following WPs:
0031-86,
Install
uninterruptable
power
supply
panel
for
communications
radio repeater Fl; 0005-87, Install
uninterruptable
panel
for
communications
radio
repeater
F2;
2139-87,
Rework conduit,
cables,
and
equipment at fire doors
485,
500,
510,
630,
642,
and
654; 3013-81, Install additional
emergency lighting units,
raceway,
supports
and cables
in Unit 3
reactor
building, control
bay
and
DG building;
and
2012-86,
Abandon cable
2ES22546-II
and
add cable
2ES202-IS2 for HPCI and
ADS separation.
During this Appendix
R review the inspector
was
informed that
the responsibility for Appendix
R
was
being
transferred
to the system engineering
group and specifically to
the Acting Mechanical
Test Supervisor.
The inspector
discussed
the Appendix
R General
Requirement
C.6 that fire detection
and
suppression
systems
shall
be designed,
installed,
maintained,
and tested
by personnel
properly qualified by experience
and
training in fire protection systems.
This item is identified as
IFI 259,
260, 296/90-05-04,
qua'lification of System Engineer to
Maintain Fire Protection
Systems.
2)
Pipe Supports
The inspector
reviewed
DCNs W4582A,
W4593A,
W4599A and
W4604A,
which involved the installation of pipe supports
in the
system.
The inspector
was able to use the
DCNs to determine
the
locations of those
supports
to
be installed
as well
as
those
supports
scheduled
to
be modified or removed.
The inspector
noted that
each
support
drawing in the
DCNs contained
the
material required
as well as the weld maps.
The
inspector
noted that
each
ECN/DCN reviewed
contained
enough
information to write WPs.
The inspector
noted
during these
reviews
and observations
that
a
large portion of the Appendix
R modifications
may not be
closed
out prior to the next Appendix
R inspection.
However,
the
majority of the modifications
should
be in the field work completed
status.
No violations or deviations
were identified in the modifications area.
4
r
e
7.
Restart Test Program
(990308)
On
March
1,
1990,
the inspector
attended
a meeting of the Joint Test
Group.
The items of discussion
were
as follows:
Approval of new JTG membership
Approval of minutes
from last meeting
Review of TE-12 and TE-13 for 2-BFN-RTP-085, Control
Rod Drive
System
Intent Change to 2-BFN-RTP-047, Turbine Generator
Control System
Review of 2-TI-186, Control
Rod Drive System
RTD/PA-085
Review of 2-TI-183, Reactor water Cleanup
System
RTP/PA-069.
No problems
were noted
by the inspector during the meeting.
8.
Action on Previous
Inspection
Findings
(92701,
92702)
a ~
(CLOSED)
URI 259, 260, 296/88-02-03,
Control of FSAR Updates
The
annual
FSAR update
had
been deficient in the past.
This
has
resulted
in
a
FSAR which cannot
be relied
upon for
purposes.
The plant
NSRB concluded that safety evaluations
required
by
must
be only partially based
upon the
FSAR with
supplemental
validation
required
by the
use
of other
licensing
documents.
The inspector
reviewed the licensee's
closure
package for
this
URI.
For
long
term corrective
action,
the
licensee
has
initiated
a
FSAR Verification
and
Update
Program.
Under this
program,
the
FSAR will be verified and updated with the results of
the Design Baseline Verification Program.
A temporary exemption
from
the requirements of 10 CFR 50.71(e)
concerning
the annual
FSAR update
was granted until July 22,
1990.
As interim controls,
the licensee
has put in place
programs
and procedures
to maintain
a
library
and
a file of
changes.
Training
was
conducted
concerning
these
issues.
The
inspector
reviewed
two training
syllabi.
Each
pointed out that the
FSAR must
be
supplemented
by
other information and that
a review of the
10 CFR 50.59 library for
information
while
performing
safety
evaluations
and
screening
reviews.
Introduction to the
SAR,
EGT121.010,
and gualified 50.57
Preparer
Training,
IGT 024.003 were reviewed.
Since
a temporary exemption
was
granted
for the
update, based
on
the controls
the
licensee
has
established,
a violation of
NRC requirements
is not warranted.
The
FSAR will be reviewed
as part of the normal
FSAR update
process
in
July 1990.
This item is closed.
15
(CLOSED)
URI 259, 260, 296/88-24-02,
High
DG Control Cabinet Internal
Temperature.
This
URI concerned
whether
a
140 degree
F temperature limit applied
to the
DG room ambient temperature
or to the inside of the control
cabinets
near
the electrical/electronic
equipment.
The
DG vendor,
Morrison-Knudson,
stated
that the
maximum allowable ambient
temper-
ature limit for the
panel
was
140
degrees
F,
and
the
maximum
localized air temperature
limit inside the cabinet
was
176 degrees
F.
A full load test
was
conducted at 2850
kw and all temperatures
were
below
140
degrees
F except for two locations.
The
vendor
recommended
an additional test,
and the shielding of one thermocouple
from radiant
heat
sources.
The additional test
was
performed
on
September
19,
1988.
MR 859966
was written to connect
the test
equipment.
The highest temperature
recorded
was
132 degrees
F, which
was
below both temperature
limits.
The inspector
reviewed the test
results,
vendor
correspondence,
and
licensee
closure
package
and
concluded that this issue is resolved.
This item is closed.
(CLOSED)
URI 260/89-20-06,
Restriction of Untrained
Personnel
From
Work Activities and
URI 260/89-20-07,
Possible
Failure to Provide
Training to gA, Radcon,
and
NE Personnel.
During
an inspection
conducted
in the area of training of licensee
personnel,
an inspector identified that site modifications engineers
had not completed
the licensee's
orientation
phase
training within
the time limit established
in the Nuclear
Performance
Plan.
This
resulted
in the
issuance
of Deviation 260/89-20-05.
The inspector
further identified that
the licensee's
training organization
had
identified various non-modifications
personnel
who had not completed
orientation
phase training or retraining within the established
time
limit.
These
URIs were
opened
pending further review of licensee
actions in this area.
The inspector
reviewed various documentation
and internal
memorandums
provided
by the licensee
during followup inspections
in this area
as
described
in IRs-89-61
and 90-03.
Additionally, NCV 260/90-03-01
was
issued for failure to correct
a
known condition adverse
to quality
related
to this
issue.
In those
followup inspections
the
NRC
determined
that the licensee
had reaffirmed their position that any
engineers,
technical staff, or managers
involved in the conduct of
actions that affect nuclear safety would receive the technical staff
and manager training.
However,
as of the close of those reporting
periods,
the licensee
had not provided the inspector with
documentation
to indicate
that
an
adequate
licensee
review
had
occurred
to verify that untrained
personnel
were restricted
from
unreviewed work.
Subsequent
to these
inspections
the inspector
was provided additional
documentation
which is related
to these
items.
The
inspector
determined
that
the
licensee
verified that for the
referenced
')
0
16
personnel
requiring training,
several
personnel
are
no
longer
employed
and
the
remainder
have either
completed
the training,
received
approved
waivers or are
scheduled
to receive
the training
during the
upcoming year.
After additional
review the inspector
determined
that
the
licensee's
quality,
technical
staff
and
management
organizations
provided
adequate
controls
to insure that
work performed
by individuals
who have not yet met the requirements
for performing unreviewed
work is reviewed
by qualified individuals.
The inspectors will continue to monitor the licensee's
activities in
this area with further review associated
with the completion of the
required training as part of the followup to Deviation (260/89-20-05).
URIs 260/89-20-06 'and 260/89-20-07
are closed.
(CLOSED)
259,
260,
296/89-18-04,
Failure
to
Provide
Cross-Disciplinary
Review of Procedures.
This violation was for failure to provide cross-disciplinary
review
of procedures
as required
by
TS section 6.8. l.l.j and
SDSP 7.4.
The
inspection
reviewed
the
licensee's
closure
package
and
sampled
several
recently revised
procedures.
SDSP 7.4 was revised to include
a comprehensive
procedure
verification review checklist.
A letter
was
issued
to all site
employees
on March 21,
1989, which discussed
the requirements
for cross disciplinary reviews
and the violation.
The inspector
sampled
several
recently revised
OIs concerning
layup
and
cross
disciplinary reviews
included
systems
engineering
and
chemistry
section.
The corrective
actions
- taken
appropriately
addressed
the issue.
This item is closed.
(CLOSED)
URI 260/89-20-08,
Corrective Action for CA(R
- This
unresolved
item questioned
the
closure
of
CARR
BFP 800695P
issued
on September
16,
1988, which identified the failure to provide
orientation training for modification engineers.
This
CARR was closed
on March 20,
1989 based
on the fact that training had
been
requested.
Closure
of
CARR
BFP 800695P
did not
meet
the
requirements
for
closure
specified
in
SDSP-3.13,
Corrective Action, Attachment
E.
This
SDSP states:
CAgRs to
be dispositioned
by providing training may be closed
when the target
audience
specified in the corrective action
has
received
the promised training.
On occasion,
closure
based
on
training of
10
percent
less
than
when
a justification is
provided
on or referenced
by the
CARR.
TVA's closure of the
CARR with greater
than
lOX of the orientation
training
incomplete
is
considered
to
be
a violation of their
procedure
SDSD-3.13.
After review by
NRC management, it has
been
decided to not issue
a
violation for premature
closure of the
CARR.
The primary reasons
are
as follows:
I
0
0
17
A deviation
from
a
commitment in the
TVA Nuclear
Performance
Plan,
Volume 3 to provide orientation the training was issued
as
Deviation 260/89-20-05.
The licensee's
September
18,
1989
response
to the deviation
coranits
to providing the
required
training
on
a
schedule
satisfactory to the
NRC.
NRC will confirm that the training has
been given in the follow-up and closure of Deviation
260/
89-20-05.
The
CA(R and its closure
have minor safety significance since,
as
documented
in TVA's response
to the
above
Deviation,
the
modifications
personnel
did not have responsibilities affecting
day-to-day
safe
plant operations
and their work was controlled
by procedures
requiring review and approval
by other qualified
individuals.
The
NRC
has
not identified
problems
that
have
occurred
due to the delay in providing orientation training.
(See
paragraph
8.c of this report.)
This item is closed.
No violations or deviations
were identified during the Followup of Open
Inspection
Items.
9.
Licensing Activities
The licensee
has frequently asked for time extensions
regarding
responses
to violations.
A review was
conducted of the licensee's
response
to
NRC
inspection
report violations,
required
by the Notice of Violation to
be
submitted within 30
days
of the
date of the letter transmitting
the
notice.
The following are
examples
of response
times that exceeded
30
days:
Re ort Number
~R
~RO
Number of Da
s
89-06
89-08
89-11
89-20
89-27
89-39
89-45
89-53
5/8/89
4/7/89
5/22/89
8/4/89
8/8/89
10/13/89
11/8/89
1/18/90
7/7/89
5/12/89
7/10/89
9/18/90
9/21/89
11/22/89
12/15/89
3/5/90
60
35
49
45
44
45
37
47
In each of the above
examples
a request for extension of the response
time
was
made
by TVA and granted
by the
NRC, however,
the extension
requests
were frequently
made
near
the
end of the
30 day period.
In the'future
the
licensee
is
requested
to notify the
NRC of
a
response
extension
request
in sufficient time,
such that the resident staff can verify that
good cause exists prior to granting
an extension.
l
18
10.
Exit Interview (30703)
The
inspection
scope
and
findings
were
summarized
on
March
16,
1990
with those
persons
indicated
in paragraph
1
above.
The
inspectors
described
the
areas
inspected
and
discussed
in detail
the
inspection
findings listed
below.
The licensee
did not identify as
proprietary
any of the material
provided to or reviewed
by the inspectors
during
this
inspection.
Dissenting
comments
were
not received
from the
licensee.
~
Item
259, 260, 296/90-05-01
259) 260, 296/90-05-02
259, 260, 296/90-05-03
259) 260, 296/90-05-04
Descri tion
IFI, ECP Corrective Actions,
paragraph
4.
VIO, Inadequate
Compensatory
Fire
Protection
Measures,
paragraph
5.a.
VIO, Failure to Follow Operating
Instruction,
paragraph
5.c.
IFI, Qualification of System Engineer
to Maintain Fire Protection
Systems,
paragraph
6.d(1).
ASOS
BFNP
CATD
CFR
CSSC
DCN
ECTG
EGT
IFI
IGT
IR
JTG
KW
kHz
LCO
Automatic Depressurization
System
Assistant Shift Operations
Supervisor
Browns Ferry Nuclear Plant
Corrective Action Tracking Document
Code of Federal
Regulations
Critical Structures
Systems
Components
Design
Change Notice
Diesel
Generator
Engineering
Change Notice
Employee Concerns
Program
Employee
Concerns
Task Group
Emergency
Equipment Cooling Water
Employee
General Training
Electrical Maintenance
Instruction
Environmental Qualification
Engineered
Safety Feature
Flow Control Valve
Fire Protection
Procedure
Final Safety Analysis Report
Inspector
Followup Item
Individual and Group Training
Inspection
Report
Joint Test Group
Kilowatt
Kilohertz
Limiting Condition for Operation
l7
I
19
LER
MAI
MMI
NE
NRC
PMI
QMDS
SDSP
SFSP
TACF
TI
TS
WP
Licensee
Event Report
Modification/Additional Instruction
Mechanical
Maintenance
Instruction
Motor Operated
Valve
Maintenance
Request
Non-Cited Violation
Nuclear Engineering
Nuclear Regulatory Commission
Nuclear Safety Review Board
Operating Instruction
Power Ascension
Primary Containment Isolation System
Plant Operations
Review Committee
Preventive
Maintenance
Plant Manager Instruction
Post Maintenance Testing
Quality Assurance
Quality Control
Qualification Maintenance
Data Sheets
Residual
Heat Removal
Revolutions
Per Minute
Reactor Protection
System
Restart
Test Program
Safety Analysis
Site Directors Standard
Practice
Spent
Fuel Storage
Pool
Surveillance Instruction
Temporary Alteration Change
Form
Test Exception
Technical
Instruction
Technical Specification
Valley Authority
Unresolved
Item
Violation
Work Plan
P
'P.
0'