ML18009A381
| ML18009A381 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 01/31/1990 |
| From: | Dance H, Shannon M, Tedrow J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18009A379 | List: |
| References | |
| 50-400-89-34, NUDOCS 9002150260 | |
| Download: ML18009A381 (21) | |
See also: IR 05000400/1989034
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report No.:
50-400/89-34
Licensee:
Carolina
Fewer
and Light Company
P.
0.
Box ".551
Raleigh,
N" 27602
Docket No.:
50-400
Facility Name:
Harris
1
License
No.:
Inspection
Conducted:
November
18,
1989
January
5,
1990
,/
Inspectors:
J'.
Tedrow, Senior Resi)ent
Inspector
'4>~ !~
.M. Shannon,
Resident
Inspector
Approved by:
C
.
H. Dance,
Se tion Chief
Division of Reactor Projects
Da'te Signed
/ ~3i ~FD
Date Signed
i)/,/
Date Signed
SUMMARY
Scope:
This routine inspect'n
was
conducted
by two resident,
inspectors
in the areas
of
plant
operation.;
radiological
controls;
security;
fire protection;
surveillance
observa:ion;
maintenance
observation;
licensee
event
reports;
non-conformance
repo";s;
system
engineer
program;
fitness for duty training;
design
changes
and modifications;
plant startup
from refueling;
followup of
onsite events;
and
1
=ensee
action
on previous inspection
items.
Results:
Two violations were identified:
Failure to perform surveillance testing for an
emergency
diesel
generator,
paragraph
3.a;
Failure
to
properly
implement surveillance
procedures,
paragraphs
3.b and 3.c.
An unresolved
item is identified in paragraph
6.a concerning
the operability of
the main
steam lines when filled with condensate.
A licensee
identified violation is discussed
in paragraph
6.b concerning
a
failure to perform
a visual inspection of check-valve internals.
9OP21-02 ", O~,OOI)4OO
90013k
0
Operational
weaknesses
are
discussed
in
paragraph ll, concerning
secondary
plant transients
causing
AFM actuations,
and paragraph
6.a, fai lure to control
heatup activities which resulted
in a main
steam safety actuation,
and filling
of main steam lines with condensate
during plant heatup.
REPORT DETAILS
Persons
Contacted
Licensee
Employees
S. Allen, Fitness for Duty Site Coordinator
D. Braund',
Supervisor,
Security
J. Collins, Manager,
Operations
- G. Forehand,
Director,
QA/AC
"C. Gibson, Director,
Programs
and Procedures
"P. Hadel, Project Spec'ialist,
Planning
"J.
Hammond,
Manager,
Onsite Nuclear Safety
"C. Hinnant, Plant General
Manager
"D. McCarthy, Unit Manager,
Site Engineering
T. Morton, Manager,
Maintenance
C. Olexi k, Supervisor, Shift Operations
"H. Powell, 'Manager, Training
"R. Richey,
Manager,
Harris Nuclear Project Department
- J. Sipp,
Manager,
Environmental
and Radiation Monitoring
- J. Smith, Supervisor,
Radwaste
Operation
"D. Tibbits, Director, Regulatory
Compliance
- R. Van Metre, Manager,
Technical
Support
- M. Wallace,
Senior Specialist,
Regulatory Compliance
E. Willett, Manager,
Outages
and Modifications
Other
licensee
employees
contacted
included
office,
operations,
engineering,
maintenance,
chemistry/radiation
and corporate
personnel.
- Attended exit interview
and initialisms
used
through-out this report
are listed in the
last paragraph.
Review of Plant Operations
(71707)
The plant
began
this
inspection
period in
a refueling
outage with the
reactor vessel
defueled.
The outage activities included repairs of damage
which
resulted
from
the
turbine
generator/main
transformer
fire
on
October
9,
core refueling,
an integrated
leak rate test of containment,
and
steam
generator
eddy-current
testing.
On
November
23
refueling
activities
(Mode 6) were
recommenced
and the core
was completely refueled
on November 25.
Following installation
and torquing of the reactor vessel
head,
the plant entered
cold shutdown
(Mode 5) at 4:30 p.m.
on December
1.
A plant
heatup
was
commenced
and
the
hot-standby
condition
(Mode
3)
reached
on
December
17.
On
December
20
a reactor start-up
was performed
followed by reactor criticality at 9:47 a.m.
Power operation
(Mode I) was
resumed
at
5: 17
a.m.
on
December
22.
The plant
continued
in
power
operation
(Mode 1) for the duration of this inspection period.
Power
was
limited to approximately
75;o due to an inoperable
main
steam safety valve
(see
paragraph
6.a for details).
The
inspector
reviewed
records
and
discussed
various
entries
with
operations
personnel
to
verify
compliance
with
the
Technical
Specificat,ions
(TS)
and
the
licensee's
administrative
procedures.
The
following records
were reviewed:
Shift Foreman's
Log; Control Operator's
Log; Auxiliary Operator'
Log;
Night, Order
Book;
Equipment
Record; Active Clearance
Log; Jumper
and Mire Removal
Log; Shift Turnover
Checklist;
and selected
Chemistry/Radiation
Protection
and
Radwaste
Logs.
Throughout
the inspection
period facility tours were conducted to observe
operations
and maintenance
activities in progress.
Some
operations
and
maintenance activity observations
were conducted
during backshifts.
Also,
during this
inspection
period,
licensee
meetings
were
attended
by the
inspectors
to observe
planning
and
management
activities.
The facility
tours
and
observations
encompassed
the
following
areas:
security
perimeter
fence;
control
room;
emergency
diesel
generator
building;
reactor
auxiliary building;
waste
processing
building;
fuel
handling
building; emergency
service water building; battery
rooms;
and electrical
.switchgear
rooms.
During these tours,
the following observations
were
made:
a.
Monitoring
Instrumentation
-
Equipment
operating
status,
area
atmospheric
and liquid radiation monitors, electrical
system
lineup,
reactor
operating
parameters,
and
auxiliary
equipment
operating
parameters
were observed
to verify that indicated
parameters
were
in
accordance
with the
TS for the current operational
mode.
During
a control
room tour
on December
30,
1989 the inspector
noted
that the
upper
and
lower power
range
nuclear
instruments
for NI-44
were
in
channel
defeat.
The
instruments
had
not
been
declared
and
no surveillance
testing
was
in
progress.
During
discussions
with
plant
operators,
the
inspector
learned
that
operators
had placed the channel
in defeat
because
of low NI readings
which caused
both
an upper
and lower flux deviation alarm.
Although
no apparent
TS action applied,
this action to remove
an
instrument
channel
from service without declaring it inoperable is considered
to
be
a poor practice.
b.
Shift Staffing -
The
inspectors
verified that
operating
shift
staffing was in accordance
with TS requirements
and that control
room
operations
were
being
conducted
in
an
orderly
and
professional
manner
.
In addition,
the
inspector
observed
shift turnovers
on
various
occasions
to verify
the
continuity
of plant
status,
operations
problems,
and
other
pertinent
plant
information during
these
turnovers.
C.
Plant
Housekeeping
Conditions - Storage
of material
and components,
and cleanliness
conditions of various
areas
throughout the facility
, were
observed
to
determine
whether
safety
and/or fire
hazards
existed.
Radiological
Protection
Program
-
Radiation
protection
control
activities
were
observed
routinely to verify that
these activities
were in conformance with the facility policies
and procedures,
and in
compliance
with
regulatory
requirements.
The
inspectors
also
reviewed
selected
RWPs to verify that the
was current
and that
the controls were adequate.
Security
Control
In the
course
of the
monthly activities,
the
inspector
included
a
review of the
licensee's
physical
security
program.
The performance
of various shifts of the security force was
observed
in the conduct of daily activities to include:
protected
and vital area
access
controls;
searching
of personnel,
packages,
and
vehicles;
badge
issuance
and retrieval;
escorting
of visitors;
patrols;
and compensatory
posts.
In -addition, the inspector
observed
the operational
status
of Closed Circuit Television
(CCTV) monitors,
the Intrusion Detection
system
in the central
and
secondary
alarm
stations,
protected
area -.lighting, protected
and vital area barrier
integrity,
and the security organization
interface
with operations
and maintenance.
Fire Protection - Fire protection activities, staffing and equipment
were
observed
to verify that fire brigade staffing
was appropriate
and that fire alarms,
extinguishing
equipment,
actuating
controls,
fire fighting equipment,
emergency
equipment,
and fire barriers
were
No violations or deviations
were identified.
3.
Surveillance Observation
(61726)
Surveillance
tests
were
observed
to verify that approved
procedures
were
being
used;
qualified
personnel
were
conducting
the tests;
tests
were
adequate
to
verify
equipment
operability;
calibrated
equipment
was
utilized;
and
TS requirements
were followed.
The following
GP-002
OST-1004
OST-1005
OST-1036
OST" 1073
OST-1804
EPT-069
tests
were observed
and/or data
reviewed:
Normal Plant Heatup
From Cold Solid to Hot Subcritical
Power
Range
Heat Balance
'ontrol
Rod and
Rod Position Indicator Exercise
Shutdown Margin Calculation
1B-SB Emergency
Diesel Generator Operability Test
RHR Remote Position Indication and Timing
Initial Criticality
EPT-152T
Main Steam Safety Valve Test at Power
EST-202
During- the performance
of-OST-1073
on December
31, the "B" Emergency
Diesel
Generator
started
but did not stabilize within the required
frequency
range.
TS 4.8.1.1.2.a.4
requires that the diesel
generator
shall
start
and
accelerate
to
synchronous
speed
with
a generator
voltage
of
6900
+690 volts
and
frequency
of
60 +1.2
hertz
(Hz)
within
a
10
second
time period.
The diesel
generator initially
stabilized at
a frequency
of
62
Hz
and
returned
to
60
Hz in
58
seconds.
Operators
declared
the diesel
inoperable at 10:40 p.m.
and
implemented
the requi rements
of TS 3.8. 1. 1 action "b".
This action
statement
specifies that with one diesel
generator
the
offsite
power
sources
shall
be
demonstrated
by verifying
correct
breaker alignment
and power availability wi,thin one hour and
at least
once per
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.
Subsequent
review of this matter
by licensee
personnel
determined
that
the
diesel
generator
was
capable
of performing its safety
function
and therefore
should
not
have
been
considered
At 9:45 a.m.
on January
1, the diesel
generator
was declared
even
though it was not capable
of satisfying
the requirements
of
TS 4.8. 1. 1.2.a.4.
implementation
of
the
requirements
of
TS 3.8. 1. 1
action "b" were discontinued at this time.
During
a
review of
the
shift
foreman's
log
on
January
2,
the
inspector
noticed
the
above
sequence
of events
and
discussed
this
situation with licensee
management.
The licensee
was
informed that
the
TS
had
not
been
complied with and that the "B" diesel
generator
should
not
be
considered
The
"B" diesel
generator
was
subsequently
determined
to be inoperable
by the licensee at 2:35 p.m.
on January
2.
Although the offsite
power
source
remained
available
during this
event, failure to continue to perform the breaker alignment
and power
availability
verification
surveillance
for
the
AC offsite
power
sources
for
an inoperable diesel
generator
every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, is contrary
to
the
requirements
of
TS 3.8. 1. 1
and
is
considered
to
be
a
violation.
Violation (89-34-01):
Failure to perform surveillance testing for an
emergency diesel
generator.
During
the
performance
of
procedure
OST-1804
on
November
28,
operators
observed
an
increase
in reactor
vessel
level.
The plant
was in
Mode
6
on "B"
RHR system cooling.
The "A" RHR system
was
aligned for standby
operation
with valves
RH-1
and
RH-2,
Reactor
Coolant
System
Loop
Suction
Valves,
open.
Procedure
OST-1804
directed
valve 1SI-322,
Refueling Water Storage
Tank Suction to "A"
RHR Pump, to be stroked
open
and then closed.
When operators
stroked
open this valve,
a flowpath was created
from the
RWST to the reactor
'vessel
through
valves
RH-1
and
RH-2,
allowing
borated
water
to
gravity drain into the reactor vessel.
The reactor
vessel filled,
and
contaminated
water
overflowed
into
the
seal
table
area
of
containment
through
open
instrument guide tubes.
Licensee
personnel
estimated
that approximately
25 gallons of water overflowed into the
seal
table
area.
No
personnel
contaminations
or equipment
damage
resulted
from this event.
Step
3. 1 of procedure
OST-1804
specifies
a prerequisite
for
the
performance
of this test
and requires that the
system being tested
be
aligned in
a
manner that will support
the
performance
of the test.
This
step
had
been
completed
and initialed by the senior control
operator.
Failure to'dequately
review the "system lineup to ensure
that the
system
was aligned for the safe
performance of the test is
contrary to the requirements
of procedure
OST-1804
and is considered
to be
a violation of TS 6.8. l.a.
Violation (89-34-02):
Failure to adhere
to the requirements
of
plant
procedures
which resulted
in
a spill of contaminated
water
into the seal
table area of containment.
Upon further review of procedure
OST-1804,
the inspector
noticed
a
discernible
lack of precautions
and limitations.
Only
a
statement
addressing
system operability and
a statement
regarding
radiological
controls
were
included.
The
lack of
any
specific
precautions
alerting- operators
to the potential
for this
type .of
event
is
considered
to be
a contributing cause for this event.
The
inspector
conducted
a
review of the
licensee's
activities to
adjust the
power
range
nuclear
instruments
following a calorimetric
calibration.
Following the calorimetric, plant personnel
noted that
the "fine" gain potentiometer
for the nuclear
instruments
did not
have
an
adequate
range
to adjust
the NI's to the required
values.
The fine gain potentiometer
is located
on the front panel of each
NI
channel
and is used
by operations
personnel
following a calorimetric
calibration to adjust
the
channel
as
needed.
Therefore,
a
course
gain potentiometer,
which is located
inside
the channel
drawer,
was
used
to
make
the calorimetric adjustment.
The inspector
discussed
this action with the
I&C foreman
and
learned that
he was not aware
that
a course
adjustment
had been
made.
Attachment VII of procedure
OST-1004 specifies
the steps
necessary
to
adjust
the
NI gain,
and
states
that if there
is insufficient
adjustment
on gain potentiometer
R303 (fine gain adjust),
then
a work
request will be initiated for I&C personnel
to adjust the course
gain
This work request
was not generated
as required.
The inspector
was also
informed that the course
adjustment
was
made
without utilizing plant
procedures.
The
existing
procedure
for
performing this evolution would not have
been
adequate
to achieve
the
required final course
adjust position.
The licensee
plans in future
cases
to set the course adjust
by detailed
work instructions
in the
work request.
The
failure
to
generate
the
work
request,
which
would alert
management
to
changing
plant conditions
as
required
by procedure
OST-1004,
is
considered
to
be
another
example
of the violation
discussed
in paragraph
3.b of this report.
'ne violation with two examples
was identified.
4.
Maintenance
Observation
(62703)
The
inspector
observed/reviewed
maintenance
activities to verify that
correct
equipment
clearances
were
in effect;
work requests
and fire
prevention
work permits,
as
required,
were
issued
and
being
followed;
quality control
personnel
were available
for inspection
activities
as
required;
and,
TS requirements
were being followed.
Various
open
and closed work request
packages
were reviewed for the safety
injection/charging
system,
in coordination
with the
system
walkdown
inspection.
No violations or deviations
were identified.
5.
Review of Licensee
Event Reports
(92700)
The following LERs were
reviewed for potential generic
impact, to detect
trends,
and to determine
whether corrective actions
appeared
appropriate.
Events that
were reported
immediately
were
reviewed
as
they occurred
to
determine if the
TS were satisfied.
LERs were reviewed in accordance
with
the current
(Closed)
LER 87-64:
This
LER reported
that
a
special
report
for
an
inoperable radiation monitor was not submitted
as required.
The inspector
reviewed
and verified the licensee'
corrective action
as
stated
in this
report.
(Closed)
LER
89-17:
This
LER
reported
an electrical
fault
on
main
generator
output bus causing plant trip and fire damage.
The licensee's
corrective action included
bus repairs,
redesign of the isolated
phase
bus
duct supports,
redesign
of the radio frequency monitor and
redesign
of
the
bus duct
fan cooling
system
to prevent debris intrusion.
All.items
have
been
completed satisfactorily
and the unit
was
returned
to
power
operation.
Review of Nonconformance
Reports
(71707)
Significant
Operational
Occurrence
Reports
(SOORs)
and
Nonconformance
Reports
(NCRs) were'eviewed
to verify the following:
TS were
complied
with, corrective actions
as identified in the reports
were accomplished
or
being pursued for completion,
generic
items were identified and reported,
and items were reported
as required
by the TS.
SOOR
89-164
reported
that
a
steam
generator
safety
valve lifted
during plant heatup.
On December
17 the plant was in Mode
3 with RCS.
pressure
at 2235
PSIG and
RCS temperature
at 557
+ 2 degrees
F.
The
"B" steam generator
PORV was isolated
due to problems with the valve
actuator.
Also, the main condenser
was under clearance
and the drain
lines from the main
steam
were closed.
Plant operators
were
experiencing
some difficulty in maintaining
the
established
temperature
band
(+2F) due to the
"B"
PORV being isolated
and
the
slow response
of the
"A" and
"C" PORVs in automatic.
The operators
'herefore
placed
the
"A"
and
"C"
PORV controllers
in
manuals
Operator
inattentiveness
allowed reactor
coolant
system
temperature
to increase
above
normal
temperature.
Due to saturation
conditions
in the
steam
generators,
the
increase
in primary system temperature
to
approximately
565
F
corresponded
with
an
increase
in
steam
generator
pressure
until
the setpoint of
a
steam
generator
safety
valve
was
reached.
After lifting the
safety
valve,
operators
promptly reduced
temperature
and pressure
which allowed the safety
valve to reseat.
The safety valve remained
open for approximately
20
seconds.
The licensee's
investigation determined that
a safety valve
for the "A" steam generator
had lifted.
The failure of the reactor operator to adequately
control the heatup
activity with the
degraded
condition of the
PORVs, is considered
to
be weak.
A lack of supervisory
involvement in the heatup
process
was
also evident.
On
December
18,
1989,
at
approximately
11:20
a.m.
a
"B"
steam
generator
safety valve (1MS-44) opened for approximately
20
seconds.
Steam
generator
pressure
had
reached
approximately
1093
PSIG which
was well below the
lowest valve lift setpoint
of
1170
PSIG.
The
setpoint for this valve had recently
been tested satisfactory during
the refueling
outage
in October.
Licensee
personnel
investigated
this
event
and
reverified
the
setpoint
for
the
valve.
On
December
21,
1989, at approximately
12:19 p.m.,
1MS-44 again lifted
for approximately
5
seconds.
Steam
generator
pressure
had
reached
approximately
1097
and
was still well below the required lift
setpoint
for
the
valve.
The
valve
was
subsequently
declared
and the requirements
of TS 3.7. 1. 1 implemented.
The main
steam line drains
had been isolated during plant heatup
and
while the
plant
was
maintained
at
hot standby.
This allowed the
steam lines to partially fill with condensate.
The water interaction
with the safety
valve could
have
been
a contributing factor in the
low lifting of the safety valves.
It was
noted
by various
plant
personnel
that
an unusually
large
amount of water
was
found
on the
building roof following the first two safety
valve lifts.
Further,
problems associated
with the turbine driven auxiliary feedwater
pump,
which
resulted
in
pump
trips
on
two
occasions,
were
attributed
to water in the
steam lines.
Also the
seismic
analysis
for the main
steam lines
may not
be valid when the
steam lines
are
filled with water.
This could lead to an unanalyzed
plant condition
during
a seismic event,
which would result in an overcooling accident
and
a containment
overpressurization
accident,
both
due to multiple
steam
line failures.
The
licensee
agreed
to
perform
a
seismic
analysis
for this
system
assuming
the
main
steam
lines
were filled
with condensate.
This item is considered
to
be unresolved
pending
completion of the licensee's
seismic analysis.
Unresolved
Item (89-34-03):
Operability of the main
steam lines when
filled with condensate.
The failure of operations
to adequately
drain the
main
steam
lines
during plant heatup,
which resulted
in
an
condition for
the turbine driven
AFM pump, potential
safety valve actuation,
and
potential
unreviewed
safety
question
is
considered
to
be
an
,operational
weakness.
b.
NCR 89-56 reported that during
a disassembly
of cold leg accumulator
check
valves
and
during
the
refueling
outage,
a
visual inspection of the valve internal
surfaces
was not performed
as
required
by
the
licensee's
inservice
inspection
program.
The
licensee's
technical
support
group
has revised its internal
document
used for reviewing
proposed
work requests
to prevent
recurrence
of
this
nonconformance.
In addition,
the valves will be disassembled
and
inspected
during
the
next
scheduled
refueling
outage.
This
matter will be
considered
to
be
a
licensee
identified
Non-Cited
Violation (NCV).
NCV (89-34-04):
Failure to perform
a visual inspection of cold leg
check valve internal
surfaces.
7.
Review of the
System Engineer
Program (71707)
A
survey
of
the
licensee's
implementation
of the
system
engineering
concept
was performed in accordance
with a
NRC Region II Memorandum dated
November
15,
1989.
The
number
of
system
engineers,
as
well
as their
responsibilities,
interfaces,
and reporting hierarchy were reviewed.
The system engineers
report directly to the Manager of Technical
Support.
Their duties
and
responsibilities
are
formally defined
in
procedure
TMM-100, Technical
Support
Conduct of Operations.
These
duties include,
in part,
log reviews,
system
walkdowns,
special test preparation,
review
of system modifications,
system description preparation,
and assistance
to
various other plant organizations.
The system engineers
perform
an active
role in troubleshooting
system
problems
and serve
as the lead contact in
coordinating corrective actions
This support provided to plant operation
is considered
to be
a strength of the licensee's
technical
support group.
No violations or deviations
were identified.
Inspection of Initial Fitness
for Duty Training (2515/104)
The
inspector
attended
selected
licensee
fitness
for
duty training
sessions
to determine
whether
required training
was
being
conducted
to
implement
the
Fitness
for Duty
Program.
A policy awareness
training
session
for
general
employees
as
well
as
a
training
session
for
supervisory
personnel
were
observed.
~ The licensee's
program
and
were
reviewed
to
determine
the
training
requirements.
The
inspector
also
discussed
the
program with the Site
Fitness
for Duty
Coordinator.
All personnel
who held
an active security
badge at the plant were required
to attend
the policy awareness
training.
Although escorts
did not receive
separate
training, their duties
and responsibilities
were discussed
during
the
policy
awareness
training.
In addition
to
the policy awareness
training,
supervisors
also
received additional
training
on their duties
and responsibilities.
The class
room discussion
did not discuss
techniques
for recognizing drugs
or indications
of
the
sale
or
possession
of drugs.
A booklet
was
distributed,
however,
which provided
information
on the identification,
use,
and description
of drug
paraphernalia.
The
booklet
was
to
be
reviewed
by the individual at his convenience.
A handout
provided
to
supervisors
included
a
copy of the
program
and
training notes
which could serve
as
a valuable future reference.
No violations or deviations
were identified.
Design
Changes
and Modifications (37828)
Installation of new or modified
systems
were reviewed to verify that the
changes
were
reviewed
and
approved
in accordance
with 10 CFR 50.59, that
the
changes
were performed
in accordance
with technically
adequate
and
approved
procedures,
that
subsequent
testing
and
test
results
met
acceptance
criteria or deviations
were
resolved
in an acceptable
manner,
and that
appropriate
drawings
and facility procedures
were
revised
as
necessary.
This review included
selected
observations
of modifications
and/or testing in progress.
'
10
The following plant change
requests
(PCRs)
were reviewed and/or associated
testing
observed:
- PCR-4007,
Electrical Penetration
Protection
- PCR-4682,
Consequences
of Losing Both S-4
Fans
PCR-4434,
480 volt Circuit Breaker
Replacement
PCR-4869,
NH Seri'es
Hydromotor Operators
- PCR-4922,
Engineering
Evaluation for Hanger
Removal
on
RHR System
i
4869
provided
instructions
to
modify various
HVAC system
actuators.
During troubleshooting
of
an
in October,
1989,
licensee
personnel
noticed that
the oil in the
actuator
motor for the
was
discolored
and
the wiring insulation
had
degraded
and
was
brittle.
The investigation
into the
cause
for this condition
revealed
that
previous
modifications
to install oil pressure
switches
and
to
eliminate
an oil control valve in the actuator,
resulted
in an increase
in
the oil
temperature
inside
the
actuator.
The
increase
in internal
temperature
resulted
in the design
temperature
being
exceeded
and thereby
reduced
the
environmentally
qualified life of
the
wiring.
These
modifications
had
been
performed
on several
actuators
in various
plant
systems.
Systems
affected
included
the
Diesel
Generator
Building HVAC, Emergency
Service Water Intake Structure
HVAC, and
Reactor
Auxiliary Building HVAC.
The licensee
has established
a schedule
to send
the affected actuators
to the vender for refurbishment.
Due to the length
of time
involved (the licensee
estimates
approximately
12-18
months to
refurbish all the actuators),
the licensee
has
performed
an evaluation
and
justification
for continued
plant
operation.
Most of the
affected
actuator s fail to the position required for the performance of its safety
function.
However,
three
actuators
associated
with
HVAC switchgear
room cooling
have
been
modified (blocked
open)
so that the
dampers will
remain
open
pending
failure.
Other
compensatory
measures
include
monitoring the Diesel Generator
Building room temperature
and placement of
portable heaters if excessively
low temperatures
are experienced.
Inspector
Followup Item (89-34-05):
Follow the licensee's
activities to
r epl ace
HVAC system actuators.
10.
Plant Startup
From Refueling (71711)
The
inspector
observed
plant activities during unit startup
following
refueling to verify that plant
systems
were properly returned to service
and that the startup
was conducted
in accordance
with approved
procedures
and in compliance with the TS.
The plant
began
deboration
to criticality on December
20 and entered
the
startup
mode
(Mode 2) at 3:23 a.m.
Criticality was achieved at 9:47 a.m.
on
December
20.
The inspectors
observed
the approach
to criticality and
ll
verified that criticality occurred within the limits calculated
by the
licensee.
- mplementation
of the following procedures
was verified:
- GP-004,
Reactor Startup
- EPT-069, Initial Criticality
No violations or deviations
were identified.
11.
Followup of Onsite
Events
(93702)
On December
20,
1989 at 8:35 p.m., during the initial start of the
a main
(NFW)
pump following the refueling
outage,
the
"A"
NFW
pump
failed to start.
The start failure satisfied
the circuitry for an
actuation
signal
due to both
MFW pumps being in a tripped condidtion.
All
AFW components
functioned
as
designed
and
plant
personnel
initiated
a
maintenance
work request
to investigate
and repair the discrepancy.
Following a breaker
check for the "A" MFW pump,
a
second attempt to start
the
pumps
was
made
on December
21, at 3:20 a.m.
The
pump again failed to
start
and
another
AFW actuation
signal
was
generated.
Subsequently,
a
third attempt
was
made to start the "A" MFW pump at 5:40 a.m.
even
though
no apparent
corrective maintenance
had
been
performed
following the last
trip.
The
"A"
MFW pump tripped
again
and
caused
another
AFW actuation.
Following this actuation,
licensee
personnel
determined that the low lube
oil pressure
switch
was
causing
the
pump to trip.
The lube oil pressure
switch was recalibrated
and the "A" NFW pump was started
successfully.
On
December
27,
1989 at 4:50 p.m., the "A" NFW pump tripped resulting in
yet another
AFW actuation.
This trip of the
"A" MFW pump
was
due to
operator
errors while operating
at
low power
(12%) with all three
feed
regulating
valves
in
manual.
Upon
increasing
flow to
two
steam
generators,
the
MFW pump recirc valve closed automatically
as
designed,
which
resulted
in
an
increase
in
flow to all
the
steam
generators.
The
operator
therefore
reduced
flow
through
the
feed
regulating
valves
in response
to increasing
steam
generator
level.
The
decreasing
feedwater flow actuated
the
MFP low flow trip and,
due to the
already
secured
"B"
MFW pump,
thereby satisfied
the
AFW actuation
logic
for loss of both
MFWs.
Once into this transient
the operator
had little
chance of recovery.
This
AFW actuation is considered
to be
an operational
weakness,
in that the
new inexperience
operator
needed
additional
guidance
during this difficult startup evolution.
12.
Licensee
Action
on
Previously Identified Inspection
Findings
(92702
92701)
(Closed)
Unresolved
Item 400/89-28-01,
Review the licensee's
engineering
evaluation of
RHR system operability with two pipe supports missing.
12
'he
inspector
reviewed
PCR 4922,
Engineering
Evaluation for Hanger
Removal
on
RHR System.
The licensee
determined that the affected
system
remained
with the
missing
supports.
Calculations
showed that
supports
adjacent
to
the
ones
missing
were still capable
of performing their
intended
function
even with the additional
loads which resulted
from the
missing supports.
(Closed)
IFI
400/89-21-02,
Emergency
Service
Water
Flow
Balance
Inconsistencies.
This
item
was
previously
discussed
in
NRC Inspection
Report
50-400/89-28.
Following
the
Licensee's
completion
of
flow
balance/pressure
testing
of the
emergency
service
water
system,
the
licensee's
nuclear
engineering
department
reviewed
the test
data
and
calculated
minimum supply flows to equipment
under worse
case conditions.
The calculations
confirmed that
the
emergency
service
water
system
was
capable of supplying adequate
cooling flow to safety related
components
as
originally designed
and
as
assumed
in
the
FSAR.
The
inspector
was
informed that
the
licensee
intended
to develop
a
computer
model
to
facilitate future
system work and to assist
in their response
to the
NRC
service water system generic letter.
13.
Exit Interview (30703)
The inspectors
met with licensee
representatives
(denoted
in paragraph
1)
at the
conclusion
of the
inspection
on
January
5,
1990.
During this
meeting,
the
inspectors
summarized
the
scope
and
findings of the
inspection
as they are detailed
in thi s report,
with particular
emphasis
on the Violations, Unresolved
Item,
and Inspector
Follow-up item addressed
below.
The licensee
representatives
acknowledged
the inspector's
comments
and did not identify as proprietary
any of the materials
provided to or
reviewed
by the inspectors
during this inspection.
Item Number
Descri tion and Reference
50-400/89-34-01
50-400/89-34-02
Violation - Failure to perform surveillance
testing
for
an
emergency
diesel
generator,
(paragraph
3.a).
Violation - Failure to adhere
to the
requirements
of plant procedures,
paragraphs
3.b
and 3.c).
50-400/89-34-03
Unresolved
Item -
Operability of the main
steam
lines
when filled with condensate,
(paragraph
6.a).
13
Item Number
(cont'd)
50-400/89-34-04
50-400/89-34-05
Acronyms and Initialisms
Oescri tion and Reference
Licensee Identified Violation - Failure to
perform
a
visual
inspection
of
cold
leg
check
valve
internal
surfaces,
(paragraph
6.b).
IFI - Follow the licensee's
activities to
replace
HVAC system actuators,
(paragraph
9).
CFR
EPT
EST
F
HZ
IFI
LER
MS
NI
NRC
OST
RHR/RH
SOOR
TMM
TS
Alternating Current
Closed Circuit Television
Code of Federal
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Engineering
Performance
Test
Engineering
Surveillance
Test
Fahrenheit
Final Safety Analysis Report
Heating, Ventilation and Air Conditioning
Hertz
General
Procedure
Instrumentation
and Control
'nspector
Follow-up Item
Licensee
Event Report
Main Feedwater
Non-conformance
Report
Non-cited Violation
Nuclear Instrumentation
Nuclear Regulatory
Commission
Operations
Surveillance
Test
Plant
Change
Request
Power Operated
Relief Valve
Pounds
per Square
Inch Gage
Quality Assurance
Quality Control
Reactor Auxiliary Building
Radiation Control Area
Residual
Heat
Removal
Radiation
Work Permit
Refueling Water Storage
Tank
Safety Injection
>>
Significant Operational
Occurrence
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