ML17311A815
| ML17311A815 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 04/07/1995 |
| From: | Wong H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17311A814 | List: |
| References | |
| 50-528-95-03, 50-528-95-3, 50-529-95-03, 50-529-95-3, 50-530-95-03, 50-530-95-3, NUDOCS 9504250069 | |
| Download: ML17311A815 (54) | |
See also: IR 05000115/2002025
Text
ENCLOSURE
U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection
Report:
50-528/95-03
50-529/95-03
50-530/95-03
Licenses:
NPF-51
Licensee:
Arizona Public Service
Company
P.O.
Box 53999
Phoenix,
Facility Name:
Palo Verde Nuclear Generating
Station,
Units 1,
2,
and
3
Inspection At;
Maricopa County, Arizona
Inspection
Conducted:
January
15 through February
25,
1995
Inspectors:
K. Johnston,
Senior Resident
Inspector
J.
Kramer,
Resident
Inspector
A. MacDougall, Resident
Inspector
B. Olson, Project Inspector
Approved:
owar
ong,
C ie
,
eactor
Jects
Branc
ate
Ins ection
Summar
Areas
Ins ected
Units
1
2
and
3
Routine,
announced
inspection of plant
status,
onsite
response
to events,
operational
safety verification,
maintenance
and surveillance
observations,
onsite engineering,
refueling
operations,
maintenance
and engineering
followup, and,licensee
event reports.
Results
Units
1
2
and
3
~
Plant
0 erations
Inspectors
observed
mixed performance
in the area of plant operations.
Unit 2
operators
demonstrated
excellent performance,
with minor exceptions,
when
performing complex
and challenging
integrated
safeguards
testing
(Section
6. I).
However, Units
1
and
3 operators
did not demonstrate
a
questioning attitude
when they noted,
but failed to investigate,
off-normal
indications
in safety, injection line pressures
(Section 3.3).
In addition,
once the concern
had
been
addressed
in Unit 1, operations
management
did not
promptly investigate
the condition in Unit 3.
9504250069 9504i9
ADQCK 05000528
8
Ll
f
I
The inspectors
observed
a Nuclear Assurance
(NA) evaluator identify that
Unit 3 operators
did not implement
a procedure prerequisite
when starting
an
essential
cooling water
pump,
and the
NA evaluator
ensured that the finding
was documented,
reviewed,
and resolved
(Section 3.2).
~
Maintenance
While maintenance
craft continued to demonstrate
excellent work practices,
some of their work instructions
were poorly developed
and required significant
adjustment
in the Field.
As
an example,
work performed
on
a Unit 2 emergency
diesel
generator
was found to be good.
However,
associated
work instructions
were found to have weaknesses
such
as conflicting torque specifications,
duplicate gasket material,
and several
pages of nonapplicable
instructions
(Section 4.2).
Similar work instruction weaknesses
were identified'y an
NA
audit,
performed in November
1994.
The audit tied work instruction
inadequacies
to weak planner qualification and training (Section 5).
The
audit report
was found to be excellent,
identifying corrective actions
along
with the findings in a succinct style.
Weaknesses
were noted in the evaluation of main feedwater
pump reliability
which followed the failure of a governor control
speed
probe (Section 4. I).
These
weaknesses
also pointed to
a failure data trending data
base
which
lacked data for important-to-safety
systems.
Maintenance
management
acknowledged
that the feedwater
pump reliability evaluation
was weak
and
continued
improvements
need to be
made to the failure data trending program.
Inspectors
found that maintenance
personnel
conducting refueling activities
demonstrated
several
strengths,
including
a very good
knowledge of the job and
the equipment.
Licensee
engineering
demonstrated
excellent technical capabilities
and
a
strong questioning attitude in their resolution of discrepancies
they
identified in an analysis
performed
by
a vendor supporting
modifications.
Engineering
management
took conservative
action to recommend
that the modifications,
which were
soon to be implemented,
be put on hold
(Section 7.1).
Inspectors
noted that longstanding
unresolved deficiencies
were receiving
greater attention.
Two issues
discussed
in the report involve plant
deficiencies
which had weak technical
reviews
and low priority for resolution.
One issue
involving essential
cooling water radiation monitors confirmed
previously identified weaknesses
in operability and safety evaluations,
timeliness of problem resolution,
and reportability (Section 7.2).
These
programs
have all received significant attention
in the past year.
In
addition, auxiliary feedwater modifications are being
implemented
which are
designed
to address
longstanding
condensation
drainage deficiencies
I
I
0'
~
~
(Section 3. 1).
Engineering
management
recognizes
that continued attention is
warranted
on these older issues.
~
Mana ement Oversi ht
Licensee
management
was responsive to weaknesses
identified in maintenance
instructions, failure data trending,
excess
use of overtime,
and engineering
issues.
Continued
assessment,
beyond the concerns
identified by the
inspectors,
was either being performed,
or was planned.
In addition,
NA was
demonstrating
a presence
in the field, asking probing questions,
and
addressing
significant issues.
Plant operators'ncorrect
assessment
of off-normal plant parameters
in the
safety injection system
and operations
management
slow resolution of the issue
indicated that continued attention
appears
warranted
in the area of operator
performance.
While engineering
management
has raised
the priority of resolving longstanding
plant deficiencies,
weaknesses
in the initial technical
reviews of
deficiencies
in the essential
cooling water system radiation monitors
and
condensate
drains indicate that continued attention is
warranted.
Summar
of Ins ection Findin s:
One unresolved
item was identified concerning
the operability and
reportability of the essential
cooling water system with nonseismically
qualified instrumentation
(Section 7.2)
~
Two violations were closed'Sections
9. 1 and
11. 1)
~
Three
Licensee
Event Reports
were closed
(Section
12).
Attachments:
1.
Persons
Contacted
and Exit Meeting
2.
1
I
)
i)f
/
I
ff
'
DETAILS
1
PLANT STATUS
1.1
Unit
1
Unit
1 began
the inspection period in a power ascension
to 100 percent
power
following the performance of a steam generator
chemistry hideout return test
and remained
at essentially
100 percent
power through the inspection period.
1.2
Unit 2
Unit
2 operated
at
100 percent
power during the first part of the inspection
period.
On February 3,
power was reduced
and the unit was taken offline in
preparation for the start of the fifth refueling outage.
On February
17,
during
a maintenance activity, startup strainers
were found in both
containment
spray
pump suction lines (discussed
in
NRC Special
Inspection
Report 529/95-08).
The unit ended the inspection
period with the core
offloaded.
1.3
Unit 3
Unit 3 began
the inspection
period at
100 percent
power and remained
at
essentially
100 percent
power through the inspection period.
2.
ONSITE RESPONSE
TO EVENTS
(93702)
2. 1
0 erations
Su
ort Buildin
Fire
Unit 3
On January
19,
1995,
a small fire broke out on the first floor of the Unit 3
operations
support building.
The fire, which started
in a battery storage
area,
was extinguished
by the automatic sprinkler system.
The storage
room
'contained
small
spare lead/acid
and nickel/cadmium batteries.
The onsite fire
department
responded
to the fire and ensured
the fire did not spread.
The
operations
support building is adjacent
to the Unit 3 control building and
houses site maintenance
and the
NRC resident inspectors'ffice.
The inspector
responded
to the control
room upon notification of the fire.
The inspector
noted that the reactor operators
were focused
on monitoring the
control board indications
and were not distracted
by the events of the fire.
The inspector
also observed site management
presence
in the control
room.
The
licensee
reviewed the emergency
plan classification
and determined that the
event did not warrant
an emergency classification
in that the fire did not
last longer than
10 minutes.
The inspector
agreed with the licensee's
determination.
The licensee
thought the fire may have
been
caused
by an employee laying
a
rubber apron across
the batteries
and causing
one of the batteries
with a
"maintenance
required tag" attached
with a fine wire to short circuit a
battery cell.
This may have
caused
the wire to get hot and start the apron
'
burning.
The inspector
concluded that the fire had
no impact
on Unit 3 plant
operations
or equipment
and the control
room personnel
responded
appropriately.
3 OPERATIONAL SAFETY VERIFICATION
(71707)
3. 1
Turbine-Driven Auxiliar
Pum
Drains Isolated
On January
25, during
a routine walkdown of the turbine-driven
AFW pump,
the
inspector
noted that the upstream
steam line drain isolation valve,
V-159,
and
the turbine trip and throttle valve below seat drain isolation valve,
V-161,
were closed.
The inspector also noted that there
was condensation
draining
out of the turbine casing drain
and the exhaust
stack drain line.
The
inspector
reasoned
that with the upstream drains isolated, all the 'condensate
in the steam
supply line had to drain through the
AFW pump turbine.
Additionally, there
was probably
some
amount of condensate
that remained
in
the upstream piping that would be forced through the turbine
when it was
started.
The inspector discussed
this configuration with the system engineer
to determine
why the drains
were isolated,
whether 'a
10 CFR 50.59 safety
evaluation
had
been
performed,
and if this configuration
had
an impact
on
pump operability.
The inspector
was informed that the drains
were isolated
in 1989 due to
a
concern with excess
flow check valves installed in the drain lines.
The
excess
flow check valves were designed
to stay
open during standby operation
to allow draining of the condensation.
When steam
was admitted to the
turbine,
they were
supposed
to close
and prevent
steam
from entering the
pump room.
The excess
flow check valves were also'sed
as the American
Society of Mechanical
Engineers
Boiler and Vessel
Code class
break between the
safety
and nonsafety
grade piping and were designed
to stop flow if the
nonsafety piping broke during
a design
bases
event.
In 1986, engineering
documented reliability problems with the excess
flow
check valves in several different systems
and
recommended
that the excess
flow
valves not
be relied
upon
as designed.
The evaluation
addressed
the impact
on
operability and the compensatory
measures
for each
system that
used the excess
flow check valves.
The evaluation
concluded that isolating the
AFW steam
supply upstream drains did not impact operability of the turbine since there
had not been
any performance
problems with turbine overspeeds
on startup
and
three of the five original drain paths
were still available.
The evaluation
did not recommend
any compensatory
measures
to periodically open the upstream
drains
and the drawings
were updated to show the isolation valves,
V-159 and
V-161,
as normally closed.
The inspector
reviewed the justification for
isolating the steam line drains
and agreed with the conclusion that the change
did not present
an operability concern.
In 1989, engineering
evaluated
the long term solution to the problem with the
AFW steam line drains.
Engineering
recommended
that the drains
be left
isolated
and administratively controlled to ensure
they remain closed.
Engineering
performed
a 50.59 evaluation
as part of this review and concluded
that the change did not create
a unreviewed safety question.
The inspector
reviewed the 50.59 evaluation
and noted that it did not include
a discussion
of the effect of the condensation
flowing through the turbine
and the
potential for turbine overspeed
events with the drains isolated.
The
inspector
noted that the initial evaluation
in l986 discussed
the potential
for turbine overspeeds
and that the turbine vendor technical
manual
did not
include any specific cautions or recommendations
for how much draining through
the turbine casing
was acceptable.
While the licensee
has experienced
overspeed trips due to condensation,
the condensation
in the turbine
was
considered
to have minimal impact.
The inspector
concluded that the decision to isolate the
AFW steam line drains
was
an example of a change
to the plant that
was performed outside the normal
design
change
process,
similar to changes
to the essential
cooling water
system discussed
in Section 7.2.
Although the drains
were isolated five years
ago,
the inspector
noted that the licensee
had not initiated actions at that
time to improve the steam line drains.
Prior to this inspection,
the licensee
had reevaluated
the steam line drains
as part of a reassessment
of AFW pump
performance
and
had initiated
a design
change to reroute the upstream drains
as part of a comprehensive
design
change to improve the reliability of the
turbine-driven
AFW pumps.
The design
change
was being
implemented
in the
current Unit 2 refueling outage
and
was
scheduled
for installation in the next
refueling outages
in Units I and 3.
The licensee
agreed that they should
have
addressed
the drainage configuration earlier.
The inspector
concluded that this issue
could have
been
addressed
earlier if
the licensee
agreed.
The inspector will review the effect of the design
change during future routine inspections.
As discussed
in Section 7.2, the
licensee
has
addressed
weaknesses
in the areas
of plant modifications,
50.59
reviews,
and engineering
follow through.
These actions
have
been
implemented
to prevent this type of issue
from recurring.
3.2
0 erator
Procedure
Usa
e
Unit 3
On February 8,'995,
the inspector
observed
control
room operators
perform
a
manual start of the essential
cooling water system.
In addition to the
inspector,
a nuclear
assurance
observer
monitored the evolution.
The
essential
cooling water
(EW) system operating
procedure
included
a
prerequisite
step that required operators
to direct
a radiation monitoring
technician to ensure that the
EW system radiation monitor was
on line.
The
reactor
operator
and control
room supervisor
noted that the monitor could not
be placed
on line based
on seismic
concerns
(see
Section 7.2).
They contacted
the radiation monitoring technician to perform sampling of the
EW system
and
continued with the procedures
The nuclear
assurance
observer
questioned
the operators
decision to continue
with the procedure without resolving the apparent
discrepancy with the
prerequisite
step
and initiated
a Condition Report/Disposition
Request
(CRDR)
to evaluate
the event.
The licensee
subsequently
revised the operating
i
0
that the change did not create
a unreviewed safety question.
The inspector
reviewed the 50.59 evaluation
and noted that it did not include
a discussion
of the effect of the condensation
flowing through the turbine
and the
potential for turbine overspeed
events with the drains isolated.
The
inspector
noted that the initial evaluation
in 1986 discussed
the potential
for turbine overspeeds
and that the turbine vendor technical
manual
did not
include
any specific cautions
or recommendations
for how much draining through
the turbine casing
was acceptable'hile
the licensee
has experienced
trips due to condensation,
the condensation
in the turbine
was
considered
to have minimal impact.
The inspector
concluded that the decision to isolate the
AFW steam line drains
was
an example of a change
to the plant that was performed outside the normal
design
change
process,
similar to changes
to the essential
cooling water
system discussed
in Section 7.2.
Although the drains
were isolated five years
ago,
the inspector
noted that the licensee
had not initiated actions at that
time to improve the steam line drains.
Prior to this inspection,
the licensee
had reevaluated
the
steam line drains
as part of a reassessment
of AFW pump
performance
and
had initiated
a design
change to reroute the upstream drains
as part of a comprehensive
design
change to improve the reliability of the
turbine-driven.AFW pumps.
The design
change
was being
implemented
in the
current Unit
2 refueling outage
and
was scheduled for installation in the next
refueling outages
in Units I and 3.
The inspector
concluded that this issue
could have
been
addressed
earlier if
the licensee
agreed.
The inspector will review the effect of the design
change during future routine inspections.
As discussed
in Section 7.2, the
licensee
has
addressed
weaknesses
in the areas
of plant modifications,
50.59
reviews,
and engineering
follow through.
These
actions
have
been
implemented
to prevent this type of issue
from recurring.
3.2
0 erator Procedure
Usa
e - Unit 3
On February 8,
1995,
the inspector
observed control
room operators
perform
a
manual start of the essential
cooling water system.
In addition to the
inspector,'
nuclear
assurance
observer
monitored the evolution.
The
essential
cooling water
(EW) system operating
procedure
included
a
prerequisite
step that required operators
to direct
a radiation monitoring
technician to ensure that the
EW system radiation monitor was
on line.
The
reactor operator
and control
room supervisor
noted that the monitor could not
be placed
on line based
on seismic
concerns
(see Section 7..2).
They contacted
the radiation monitoring technician to perform sampling of the
EW system
and
continued with the procedure.
The nuclear
assurance
observer
questioned
the operators
decision to continue
with the procedure without resolving the apparent
discrepancy with the
prerequisite
step
and initiated
a Condition Report/Disposition
Request
(CRDR)
to evaluate
the event.
The licensee
subsequently
revised the operating
,
i
I
i
procedures
in all three units to direct the radiation monitoring technician to
perform specific actions if the radiation monitor could not to be placed in
service.
The inspector
noted that the
EW radiation monitors
had
been
removed
from
service
in April 1994,
and concluded that control
room personnel
in all three
units
had
been "operating
around" the prerequisite
step in the operating
procedure,
demonstrating
a lack of ownership of the procedure.
The inspector
noted the persistence
of the nuclear
assurance
observer to correct the
deficiency.
The inspector will evaluate
the final
CRDR recommendations
as
part of a routine inspection.
3.3
Safet
In 'ection
Tank Drain and Fill Isolation Valve Leaka
e
On February
9, during
a routine tour of the Unit
1 control
room, the inspector
noted that the "B" train hot leg SI check valve leakage
pressure
indicator,
(PI) 391,
was reading
about
600 psi.
This gage is used to indicate
check valve leakage
from the reactor coolant
system hot leg through the
shutdown cooling suction line into the high pressure
safety injection (HPSI)
hot leg recirculation line.
The inspector
asked
the operators
why there
was
a
difference
between
the "A" train check valve leakage indicator PI-390, which
was reading less
than
50 psi,
and the "B" train indicator PI-391,
which was
reading close to safety injection tank (SIT) pressure.
The operators
told the inspector that running the
pump during routine
surveillance tests
caused
pressure
to become
locked
between
the two hot leg
injection check valves.
The inspector
had previously questioned
operators
on
the difference
between
the readings of PI-390
and PI-391
and was given the
same explanation.
The inspector
asked
the operators
when the last
HPSI test
was performed
and
why PI-391
was reading the
same pressure
as
SIT pressure.
As
a result of the inspector's
questions,
operators
investigated
the situation
and determined that the
2B SIT drain
and fill isolation Valve SIB-641,
and the
HPSI fill isolation Valve SIB-322 were leaking past their seats.
The Unit
1 shift supervisor initiated work requests
for the leaking valves
and
called the system engineer to discuss operability of SI Valves
641
and 322.
The system engineer
stated that these
valves were not required to be zero
leakage
valves
and that the small leakage did not affect operability of the
SIT system.
The shift supervisor told the inspector that
an operability
determination
(OD) would be performed.
The inspector
agreed with the shift
supervisor's initial determination that the leaking valves did not create
an
immediate safety concern.
The inspector
noted similar differences
between
the pressure
readings for
PI-390
and PI-391 in Unit 3.
On February
14, the inspector
asked
the Unit 3
shift supervisor if he was
aware of the leaking
SI valve problem in Unit
1 and
evaluated
the condition in Unit 3.
The Unit 3 shift sunervisor
was not aware
of the problem
and
had not conducted
a review to determine
the extent of
leakage
in Unit 3.
(
The inspector
was informed that the Unit
1 shift supervisor
had sent the
information concerning
the leaking SI valves to the Unit 3 control
room on
February
10.
As
a result of the inspector's
questions,
the information from
Unit
1 was again provided to the Unit 3 control
room and
a troubleshooting
plan was developed
to determine
the source of the leakage
in Unit 3.
On
February
16, Unit 3 operators
gathered
information and determined that they
also
had
a leaking SIT drain
and fill isolation valve and
a leaking
HPSI fill
isolation valve.
On February
22, the inspector
reviewed the results of the test
conducted
in
Unit 3 and noticed that
an
OD for the leaking valves
had not been
performed.
The inspector
noted that although the SIT drain
and fill isolation valve
and
HPSI fill isolation valve were not required to be zero leakage
valves,
they
did receive
a safety injection actuation
signal to shut to maintain SIT
inventory in the event of a safety injection actuation
signal during SIT fill
or drain operations.
The inspector
questioned
the shift supervisor
about the
impact
on SIT integrity with the leaking valves
and whether
an
OD was
appropriate.
The shift supervisor told the inspector that the test data
had
been
forwarded to engineering
and that they did not think an
OD was required.
The inspector
then
informed the operations
department
leader
who agreed with
the inspector, that
an 00 was required.
On February
22, the system engineer
completed
a 'draft evaluation of the
condition in both Units
1
and 3,
The system engineer
determined that the SIT
system
was operable
because
the leakage,
in both cases,
was very small
(less
than 0. 1 gpm)
and the isolation valves
between
the SIT drain
and fill header
and the'reactor
drain tank and the refueling water tank were leak tight.
As
a
result,
the level in the SITs would stay constant
once the drain header
was
pressurized
through the leaking valves.
Additionally, system
alarms
would
alert the operators
to an increase
in boundary valve leakage if the condition
degraded.
The inspector
reviewed the design basis
manual,
the Updated Final
Safety Analysis Report,
and operating
procedures
.and
agreed with the
conclusion that the existing condition in Units
1 and
3 did not impact SIT
operability.
The inspector also noted that the system engineer
conducted
a
detailed,
thorough operability evaluation.
The inspector
had the following conclusions
concerning operator
performance
in
Units
1
and 3:
~
Control
room operators
did not question
an indicated discrepancy
in
control
board indication for safety-related
plant equipment.
The
operators
had developed
an assumption
which was not based
on fact.
While the condition was eventually determined to not be
safety-significant,
a more significant condition could have provided
similar indication.
Operations
management
did not aggressively
pursue identifying and
evaluating the condition in Unit 3.
l
,
At the exit meeting,
the Director of Operations
stated that operator
performance
did not meet his expectation
and that
he would reemphasize
the
importance of understanding
plant status,
particularly control
board
indications.
The inspector
concluded that these
actions
were appropriate.
4
MAINTENANCE OBSERVATIONS
(62703)
4. 1
Main Feedwater
Pum
S eed
Probe Failure
Unit
1
On January
21,
1995,
Speed
Probe
2 on the Unit
1
MFW Pump
A began to fail
intermittently,
The
Speed
Probe
2 failure did not impact the operation of the
HFW pump control
system,
since it acts
as
a backup input and is relied upon
only in the event of the failure of Speed
Probe
1.
However, the additional
failure of Speed
Probe
1 would result in the
MFW pump turbine governor valve
opening fully and causing
the
To repair the failed
speed
probe,
the
MFW pump would have to be removed
from service
and
power
reduced
to approximately
60 percent
power.
Plant operations
management,
with support
from maintenance
engineering,
decided to place the repair of the failed speed
probe
on hold for 10 weeks
until the start of the Unit
1 refueling outage
schedule.
The decision
involved
an evaluation of the risks of a power reduction late in core life,
the failure history of HFW speed
probes,
and the relatively short interval to
the outage.
The inspector
reviewed the failure history of the
speed
probes.
The data
was
provided
by maintenance
engineering with the support of system engineering,
and the maintenance
failure data trending group.
In general,
there
had
been
a
significant number of speed
probe failures in all three units.
However, the
inspector
found that the
MFW speed
probe failure history provided
by the
licensee
was inaccurate
and incomplete.
Although the data included several
HFW speed
probe failures,
the maintenance
engineer
noted that several
additional
speed
probe failures
had not been
picked
up in their data
search.
The inspector determined that analyses
to determine
a cause for the
speed
probe failures
had not been performed.
Additionally, the licensee
had not
identified
a mean time for speed
probe failure and noted that performing this
type of analysis
would be difficult.
The inspector
found that maintenance
had
not determined
the
age of the remaining
speed
probe in the Unit
1 "A" MFW pump
and considered
that this should
have
been
a key input to an analysis of MFW
pump reliability.
Maintenance
engineering
subsequently
determined that the
speed
probe
had
been installed in 1993.
The inspector discussed
the quality of the data with maintenance
failure data
trending
(FDT) personnel.
The licensee
recognized that
FDT historical
information, especially in balance of plant systems,
was weak.
In the case of
the
HFW speed
probes,
the
FDT program listed only three failures.
While a
significantly greater
number of failures were found in the maintenance
database,
these failures were listed under at least three different component
'
-10-
identifications.
Additionally, the maintenance
database
did not consistently
provide
an apparent
cause of failure.
The inspector discussed
the weaknesses
in the data provided to plant
operations
and the weaknesses
in the
FDT program with the Director of
Maintenance.
He concurred that the analysis
provided to plant operations
had
not met his expectations
and discussed it with his staff.
Additionally, he
recognized
the weaknesses
in the
FDT program
and the role that it would play
in the licensee's
implementation of the
10 CFR 50.65 maintenance
rule.
The
inspector
noted that
a weakness
in the
FDT program
had
been identified by the
licensee
previously
and
a maintenance
Level
1 action
had
been initiated to
improve the acquisition of data into FDT.
The inspector
found these
actions
to be appropriate.
4.2
Emer enc
Diesel
Generator
Work Order Deficiencies
Unit 2
On January
25, the inspector
observed
mechanical
maintenance
personnel
performing corrective maintenance
on the jacket water heater
pump
and the
prelube oil circulation
pump
on
EDG A in Unit 2.
Both of these
pumps
had
leaks
from the mechanical
seal
which were being replaced
as part of a
scheduled
online outage of the
EDG.
4.2. 1
Jacket
Water
(JW) Heater
Pump
The licensee
uses
model maintenance
instructions
(NI) for repetitive work to
establish
consistency
and to reduce
the
amount of prejob planning for routine
work.
The inspector
noted that the work order for reassembling
the
JW heater
pump was
an HI that
had
a step to torque the fasteners
using the general
torque specification
from the
EDG vendor technical
manual.
These
specifications
were" used if specific torque requirements
were not available.
However, the mechanic noticed that the section of the vendor technical
manual
for the
JW heater
pump included instructions to torque the fasteners
using
good mechanical
judgment.
Therefore,
the HI provided confusing guidance
on
the torque requirements
for the fasteners,
The mechanic
discussed
the situation with the
NA evaluator observing
the job
and decided to use the instructions
which specifically applied to the
heater
pump (tighten the fasteners
using good mechanical
judgment).
The
NA
evaluator also noted that the HI specified
two different class
and item
numbers for the
pump suction gasket.
The
NA evaluator
suggested
that the
mechanic
discuss this problem with the
shop planner.
The inspector
observed
a questioning attitude
by the mechanic
and the
NA
evaluator
to find and solve the problem with the HI.
In spite of the
instruction deficiencies,
the mechanic
was able to eventually perform the work
appropriately.
However,
the inspector
was concerned
that these technical
errcrs
were not identified prior to sending
the paperwork to the field.
The
inspector discussed
this observation with the mechanical
maintenance
section
leader.
The inspector
asked
the section leader if he expected
the planner to
conduct
a technical
review of NIs prior to including it in a work package
sent
I
0
-11-
to the field.
The section leader stated that the planners
were tasked with .
reviewing the MIs when they
had time, but were not expected
to review each
package prior to sending
the work to the field.
As
a result,
the planners
were relying on feedback
from the field to correct technical
problems with the
work order.
The section leader
agreed
that reviewing the NIs needed
to be
a priority and
asked
the planner assigned
to his section to conduct
a technical
review of the
ten most used Hls.
In addition,
the planner
changed
the HI to reflect the
correct torque value
and gasket
class
and item number.
The inspector
noted
that weaknesses
in MI had
been identified by licensee
organizations
and
improvements
were in progress.
The inspector
concluded that the licensee's
initial corrective actions
were appropriate.
4.2.2
Pre-Lube Oil Circulation
Pump
The inspector
observed
the mechanics
performing work on the prelube oil
circulation
pump for
EDG A.
The inspector
reviewed the work order
and noted
that the objective of the work order
was to remove,
rebuild,
and replace
the
prelube oil circulation
pump using
a model
MI.
Because
the inspector
had
difficulty in following the HI, the inspector
asked for clarification of the
NI.
The mechanics
informed the inspector that they were only replacing the
pump mechanical
seal
and that most of the instructions
were unnecessary.
The inspector
observed
the remaining work on the
pump
and did not have
any
concerns
with the performance of the work in the field.
The inspector
noted
that the mechanics
were able to determine
the appropriate
steps
to be
performed.
However,
the inspector
was concerned
that the workers were
provided with a large
amount of unnecessary
instructions for the job they were
performing.
The inspector
noted that the mechanics
in the field had to mark
out eight of the ten
pages of work order instructions.
The inspector
was
concerned
that having
a complex work document
where
a simple document 'would
suffice may reinforce the tendency of the workers to not refer to the
instructions,
increasing
the chances
of a mistake.
The inspector
asked
the responsible
section leader
and department
leader if
the planner
should
have lined out the unused
steps prior to sending
the work
order to the field.
Neither individual expected
the planners
to remove
unnecessary
steps prior to sending
the work order to the field.
They felt
this type of review was too time consuming for the planners
and that the
workers in the field were able to do the review.
The inspector
reviewed the
work order
and noted that it would only have taken
about
5 minutes for the
planner to review the work order
and
remove the unnecessary
steps.
The inspector discussed
this issue with the Director of Maintenance
and
learned that it was his expectation that planners
delete
unnecessary
work
order instructions prior to sending
the package
to the field.
The licensee
had
a long term action to "sectionalize"
MIs so that only the applicable
portion of the NI would be produced for the work package.
The licensee
also
created
a
new NI for replacement
of the
pump mechanical
seals.
!
1
0
-12-
The inspector
concluded that each level of supervision
in the mechanical
maintenance
department,
from the team leader through the director of
maintenance,
had different expectations
for the level of prejob work order
review required.
The inspector
has identified previous
weaknesses
in the
quality of work instructions.
In addition,
an
NA audit found weaknesses
in
work instructions
and found problems
in planner qualification
and training.
At the exit meeting,
the Director of Maintenance
informed the inspector of an
effort by the maintenance
support group to improve planner qualification
training.
Maintenance
support also identified the problem with inconsistent
expectations
concerning
the responsibilities of planners.
At the
end of the
inspection
period maintenance
support
had completed its review but was still
developing
how to incorporate
management's
expectations
into the revised
training program.
The inspector will review the revised planner
qualifications
and the quality of work orders during future routine
inspections.
4.3
Motor-0 crated
Valve
Maintenance
and Testin
on Valve SIA-HV-696
Unit 2
On January
18,
1995,
the inspector
observed
portions of routine
maintenance
and testing
on shutdown cooling heat
exchanger outlet Valve SIA-
HV-696.
The inspector
noted that to perform the testing,
the valve motor was
operated
frequently
by technician.
During the testing,
the inspector
observed
that the motor casing
became
hot to the touch.
The valve took approximately
90 seconds
to travel its full stroke.
Additionally, in some instances,
the
valve motor was momentarily energized
or "jogged" several
times in succession
to position the valve for specific tests.
The inspector
was concerned
that
motor heating could lead to long-term motor winding degradation.
The inspector
asked
the
MOV technician if guidance
had
been provided
concerning
the potential for overheating
valve motors during testing.
The
technician
responded
that,
in general,
they feel the casing during testing
and
take
a break from testing
when the casing gets hot.
The inspector questioned
whether this general
guidance
was consistent
with
vendor recommendations.
The
MOV maintenance
engineering
section leader
showed
the inspector
a letter from the vendor,
Limitorque, to another utility
concerning jogging capabilities
as well as full load run times.
The guidance
provided
by the vendor
stated that the valve operator could
sustain
locked rotor current for up to
10 seconds.
The vendor noted that
a
motor starting event would produce
locked rotor current for less
than
1 second,
which would allow ten consecutive starts.
To account for variables,
the vendor
recommended
that these
motors
be limited to five consecutive
starts
with a cooling period of an hour or more between restarts.
The inspector
noted that this guidance
was more conservative
than the practice
observed
in the field, when the inspector
noted that in the span of
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the
valve was fully stroked six times
and then jogged
up to six consecutive
times
I
,
j
)
-13-
at least twice.
The inspector discussed
this with the
MOV maintenance
engineering
supervisor.
He had reviewed the motor starting current traces
and
determined that Valve SIA-HV-696 achieved
locked rotor current for
0.25 seconds
each
time the valve was cycled.
According to vendor guidance,
this would allow up to 40 consecutive starts.
Based
on this data,
the section
leader
concluded that the motor had not been
damaged.
The inspector
found the
review to be acceptable.
The
MOV maintenance
engineering
section leader stated that additional training
covering the vendor recommendations
was provided to
MOV maintenance
technicians.
The training covered
the vendor's
recommendations,
and while it
did not endorse
the conservative
five start limit, it recommended
that caution
be applied.
The inspector
noted that the vendor's
guidance
was provided in September
1994,
and did not appear
to have
been provided to all licensees.
The
maintenance
engineering
section leader stated that while case-by-case
recommendations
had
been provided,
the vendor
had not provided consistent
generic guidance
to all facilities.
The section leader stated that
he would
pursue developing appropriate
generic
g'uidance for the evaluation of the
users
group.
These
recommendations
would then
be passed
to the vendor for
implementation.
The inspector
found these
actions to be appropriate
and will
follow the licensee's
actions during routine inspection.
4.4
Contamination
on Valve Stem
Unit 2
During
a routine tour of outage
work in the Unit 2 auxiliary building, the
inspector
noted
some boric acid contamination
on the valve stem of the "8" low
pressure
pump miniflow valve.
The inspector
noted that the valve stem area
was not identified as
a contaminated
area.
The inspector
asked
the
technicians
working on the valve actuator if they informed
a radiological
protection
(RP) technician that the contaminated
area
needed
to be marked.
The technicians
stated that they were
aware of the contamination,
but did not
think it was necessary
to inform the
RP technician
and
have the area
marked.
The inspector
informed
an
RP technician of the situation
and the, contamination
was surveyed
and properly marked.
The inspector
concluded that although the technicians
were using good
practices
to keep themselves
from getting contaminated,
they were not
sensitive of the
need to identify the contamination
to minimize the potential
for other workers to get contaminated.
The inspector discussed
this
observation with the valve services
section leader
who agreed that the
technicians
should
have
been sensitive to the broader
issue of identifying and
marking contaminated
areas.
The section leader stated that this observation
would be discussed
during industry events training.
The inspector
concluded
that these
actions
were appropriate.
4,
~
l
4.5
EDG Ins ection
Unit 2
On February
7, the inspector
observed
engineering
perform boroscope
inspections
of the Unit 2
8 as part of the
18 month
EDG inspection.
Based
on these
inspections,
engineers
recommended
that seven cylinder heads
be
removed
so that the cylinder liners
and pistons
could
be inspected.
The inspector
observed
the mechanical
maintenance
EDG team remove the cylinder
heads,
replace
several
liners
and pistons,
and reassemble
the affected
components.
The inspector
observed
excellent work practices
by the
EDG team
in the area of foreign material
exclusion control, identification and control
of removed parts,
cleanliness,
and work package
usage.
The inspector
also
observed
good communications
and
team work between
the maintenance
and
engineering
groups.
4.6
Other Maintenance
Observations
~
,
Uninterruptible
Power Supply for Emergency Lights for Safe
Shutdown
Monthly
Unit 3
~
Nitrogen Supply Valve Replacement
for Steam Generator
Atmospheric
Dump
Valve Control
Unit 3
t
~
MOV Actuator Gear Replacement
- Unit 2
5
EFFECTIVENESS
OF LICENSEE EQUALITY ASSURANCE
(40500)
The inspector
reviewed
an audit performed
by the
NA - Maintenance
organization
of maintenance activities.
The audit was performed in November
1994.
The
inspector
noted that the audit identified weaknesses
in the quality of MI and
the training provided to those
who write the instructions,
The audit included
corrective actions to address
these
weaknesses.
As noted in this
and recent
inspection reports,
inspectors
identified similar weaknesses
in the quality of
MI.
The audit team also identified weaknesses
in the implementation of the
preventive maintenance
basis
program
and in the identification, evaluation,
and documentation
of 10 CFR Part 21 reportability and root cause of failure
analysis
requirements.
The audit also
assessed
the significance of the
findings and outlined corrective actions.
The inspector determined
the licensee's
audit was performed
by
a multi-
discipline team,
including individuals from the maintenance
organization.
The
inspector
noted that the audit report represented
a change
in format from
previous audit reports
in that it was focused
on inspection findings
and
captured
planned corrective actions.
While the overall length of the report
was substantially shorter
than previous audits, primarily since it did not
j
I
-15-
capture audit planning detail,
the inspector
found the audit findings were
presented
in an easy to follow format, allowing the reader to focus
on the
results.
6
SURVEILLANCE OBSERVATION
(61726)
6. 1
Inte rated
Safe
uards Testin
Unit 2
On February
6,
1995,
the inspector
observed
sections of Surveillance
Procedure
Class
lE Diesel
Generator
and Integrated
Safeguards
Surveillance
Test
Train B.
The inspector
noted that the test director
demonstrated
strong
knowledge of the test procedure
and expected
plant
response,
and that the control
room supervisor maintained
strong
command
and
control of the unit.
The inspector
noted
good communications
between
engineering,
operations,
and other members of the test
team during the
performance of the test.
The inspector
noted
some minor weaknesses.
Although independent
verifications
had
been
performed,
they had not been adequately
documented.
In addition, the
inspector
observed
several
instances
during the restoration of the plant
equipment
at the conclusion of the test where the operators
would manipulate
plant equipment,
but not announce
the expected
alarms to the crew before the
alarms occurred'he
inspector notified the shift supervisor of the
weaknesses
and the shift supervisor
subsequently
addressed
the weaknesses
to
the operating
crew during the next preevolution brief for safeguards
testing.
The inspector
concluded that the shift supervisor's
corrective actions
were
adequate
and that the overall performance of the integrated
safeguards
testing
was good.
6.2
Other Surveillance
Observations
~
Emergency
Safeguards
Features
Actuation System
Subgroup
Relay Testing
Unit
1
7
PLANT ENGINEERING
(37551)
7.1
Steam Generator Modifications Put
on Hold
On February
13,
1995,
the licensee
decided that modifications planned
on the
Unit 2 steam generators
would not be performed
as
scheduled
during the Unit 2
refueling outage.
Preparations
for the modifications, involving the extension
of the
feed rings
and divider plate,
were underway
when the
modification was put on hold.
The licensee
made the decision to stop the
modification after they had identified weaknesses
in the modification
calculations
performed
by the vendor,
Combustion
Engineering.
The licensee
had found that the vendor
had not accurately
modeled the steam generator
geometry.
j
-16-
The inspector determined
the licensee
engineering
had aggressively
pursued
questions
developed
during their review of the modifications.
When answers
to
these
questions
demonstrated
weaknesses
in the vendor's analysis,
the licensee
lost confidence that
a definitive resolution could
be reached
by the
end of
the Unit
2 outage.
The inspector
noted the licensee's
decision to put the
modification on hold was conservative
and demonstrated
good quality
engineering.
As
a result of the weaknesses
in the vendor's
analyses
supporting the steam
generator modifications,
the licensee
plans to perform
an audit of the vendor.
In addition,
the licensee
put
on hold plans to install the
modifications in the Unit I refueling outage,
scheduled
to begin
on April 1.
7.2
EW Radiation Monitors
As discussed
in Section 3.2 of this report,
in April l994, the licensee
had
taken'he
process
radiation monitors
(RU
2 and 3) for the
EW system out of
service
by closing the inlet and outlet isolation valves to establish
a
seismic/nonseismic
code break.
The inspector
reviewed this decision
and other
related activities concerning
EW system configuration control.
The inspector
found significant weaknesses
in the historical control of
EW system
configuration
as
we11
as
weakness
in how the
EW system configuration
was
controlled at the time of the inspection.
7.2. I
Ori inal
S stem Desi
n
The
EW system
includes
two closed cooling trains which are cooled
by the
ultimate heat sink spray
pond system,
The
EW system provides safety-related
cooling to the shutdown heat
exchangers
and the essential
chilled water heat
exchangers.
It also provides
backup cooling to other nonvital radioactive
systems
such
as spent fuel pool cooling
and the reactor coolant
pumps.
Normal
cooling to those
systems,
supplied
by the nuclear cooling water
system,
is
isolated
from the
EW system in an event.
Each
EW train was provided with a surge tank to allow for changes
in pressure.
Level transmitters
on the surge
tanks were provided to control solenoid
operated
valves
For makeup to the
EW system from the preferred
nonsafety-related
demineralized
water
and the backup safety-related
condensate
transfer
(CT) system.
Each train of the
CT system
was designed
to provide
approximately
50
gpm of makeup
from the condensate
storage
tank
(CST).
The
system also
was designed
to provide makeup to the
JW cooling surge tank
and to the essential
chiller cooling system.
The
EW radiation monitors,
RU
2 and
RU 3 were installed in the
EW system to
meet
10 CFR Part 50, Appendix A, General
Design Criterion 44, cooling water
system requirements
for leakage detection.
The Updated
Final Safety Analysis
Report,
Section 9.2.2,
stated that to detect
leakage,
high and low level
alarms for the surge tank
and the radiation monitors provided leakage
detection.
The NRC's Safety Evaluation Report stated that each train of
EW
'
I
)
0
-17-
was provided with a "continually operating radiation monitor which alarms in
the control
room
.
RU
2 and
RU
3 were initially installed
as nonseismically qualified.
Flow
orifices in the inlet and outlets of the
RUs were used to reduce
EW system
leakage
in the event the
RUs failed following a seismic event.
A calculation
performed in 1984, prior to plant operations,
demonstrated
that worst case
leakage
through the
RUs would be
on the order of 12.5
gpm and, therefore,
within the capacity of CT makeup.
A similar calculation
had also
been
performed to demonstrate
that the failure oF nonseismic
pump discharge
pressure
instruments
would also
be within the capacity of CT system
makeup.
7.2.2
Historical Design
Changes
The inspector
found that the licensee
did not recognize,
until 1993, that the
operability oF the
EW system
depended
on the operability of the
CT system.
During this period,
the
CT system
was
removed
from service
and the
EW system
was not declared
Additionally, in 1992,
the licensee
disabled
the
ability of the
CT system to provide automatic
makeup to the
EW system.
According to the design
basis for the
EW system at the time, both trains of
EW
could have
become
inoperable within 7 minutes of a seismic event.
The licensee
then developed
a preliminary evaluation that the
EW system
would
have
remained intact following a seismic
event
and was not inoperable with CT
makeup unavailable.
However,
some
issues
are still unresolved.
Based
on the
significance of the issue,
the inspector
performed
a thorough review of the
design
change history for the
EW and
CT system
and noted weaknesses
in
operability and safety evaluations,
timely resolution of engineering
issues,
and reportability evaluations.
The following is
a summary of the significant evaluations
and resulting
system
changes
since
1992
on the
EW and
CT systems:
~
In 1992 the licensee
documented
a concern
in
CRDR 9-2-0122 regarding the
indicated level in the
CST versus
the minimum level referenced
in
Technical Specification (TS) 3.7. 1.3.
TS 3.7. 1.3 required that
level
be
"
.
.
. at least
25 feet
(300,000 gallons)."
The licensee
determined that due to instrument
loop accuracies
and errors in the
assumption for the lower instrument tap,
a level of 29.5 feet
was
needed
to assure
a volume of 300,000 gallons.
The licensee
determined that
levels
had
been routinely maintained
above 29.5 feet.
As corrective
actions,
administrative limits were placed to maintain
CST level
above
29.5 feet.
To resolve
CRDR 9-2-0122
and to reserve
CST volume for decay heat
removal,
the licensee
evaluated
the demands
placed
on
CST volume by
systems
supplied
by the
CT system.
In the review of the
EW system,
the
licensee
focused
on
a
1988 concern
regarding unreliable
excess
flow
(
l
l
l
I
-18-
used
as the seismic/nonseismic
code break point for
EW
surge
tank level instruments.
The review did not address
either the
system radiation monitors or the
pump discharge
pressure
instruments.
The evaluation
concluded that,
even with failed excess
flow check
valves,
the
EW system would not lose sufficient volume to impact
EW pump
net positive suction
head before operators
could take mitigating action.
In
a July
1992 resolution,
to preserve
CST volume, the licensee
decided
to isolate
CT makeup to the
EW system
by closing
a manual
valve.
In Hay 1993,
as part of the design
basis
review project,
the licensee
identified that the
RUs
and
pump discharge
PIs,
which were both not
seismically qualified, would provide
a worst case
leakage of 31.5
gpm
following a seismic event.
The immediate disposition of the evaluation
of this issue
CRDR 9-3-0422
was to isolate the
RUs
and PIs.
However,
within a few weeks,
engineering
reopened
the isolation to the
RUs based
on engineering
findings that the
RUs could probably
be seismically
qualified and that there
was
a low probability of an earthquake
in the
six months estimated for resolution.
The PIs remained isolated
and are
opened
only to support testing.
In April 1994,
the licensee initiated
CRDR 9-4-0271 to address
the
unresolved
seismic qualifications identified in
CRDR 9-3-0422.'ne
of
the immediate actions of the
CRDR was to isolate
RU
2 and
RU 3.
A
subsequent
review was initiated to either resolve the seismic
qualification of the
RUs or to abandon
them in place.
As compensatory
measures,
the licensee initiated routine sampling of the
EW system.
Disabling the
RUs was performed in accordance
with a
10 CFR 50.59 safety
evaluation.
The inspector
noted the following weaknesses
during this review of these
problem evaluations
and resolutions.
7.2.2.1
Operability and Safety Evaluation
Weaknesses
The inspector identified the following weaknesses
in the operability and
safety evaluations
that were performed:
k
The disposition of CRDR 9-2-0122 to isolate the automatic
CT makeup
failed to account for the original
EW system design
basis
which required
CT makeup.
Specifically, the evaluation did not consider the nonseismic
RUs
and PIs.
The decision to isolate automatic
CT makeup
was
a poor resolution for
the problem as identified in
CRDR 9-2-0122,
regardless
of engineering's
failure to consider
the nonseismic
RUs and PIs.
Since the calculations
demonstrated
that
CT makeup to the
EW system would have
had
no effect on
EW system operability and little impact
on
CST volume,
a better
resolution would have
been to leave the system aligned "as-is,"
'
f
-19-
The immediate disposition of CRDR 9-4-0271
was to isolate
RU 2 and
RU 3.
The change
was supported
by
a
10 CFR 50.59 evaluation
(94-0091).
The
evaluation
noted that manual
sampling would be initiated to detect
leakage of radioactive fluids into the
EW system.
However,
the
inspector
noted that the evaluation did not specifically address
the
replacement
of an automatic
system with a manual
system.
The inspector
considered
that
a thorough evaluation
would have
addressed
the frequency
of sampling
needed
to compensate
for an automatic
system
and the
availability of other leakage detection,
such
as the
EW surge tank level
instruments.
7.2.2.2
Operability Evaluations
At the
end of the inspection period,
the licensee
had not completed
an
evaluation to determine
whether the system would have
remained intact.
While
it appeared
that the
RUs could
be qualified in the "as-is" configuration,
modifications
had
been
performed to the PIs to be considered
seismically
qualified.
This would indicate that the PIs could have failed following a
seismic event.
The licensee,
during their reviews of both
CRDRs 9-3-0422
and 9-4-0271,
identified that the issue
was "potentially" reportable.
In a corrective
actions audit,
performed in December
1994,
NA identified that this was
one of
three
examples of issues
which had not received appropriate
followup
reportability reviews.
They noted that there
was
no tracking mechanism to
ensure that once the engineering
evaluation
was performed,
a final
reportability evaluation
was performed.
An action item was subsequently
opened
and
was being tracked.
The inspector
noted that,
as
a result of the
NA audit,
and findings associated
with a special
inspection
(NRC Inspection
Report 50-528/95-08;
50-529/95-08;
50-530/95-08),
the licensee
has initiated actions to improve reportability
training
and
enhance
the evaluation
process.
The inspector
considered
the
operability evaluation of this issue to be
an Unresolved
Item (530/9503-01).
7.2.2.3
Slow Resolution of Technical
Issues
CRDR 9-3-0422
was initiated May 1993,
and identified that the
RUs need to
either
be seismically qualified or abandoned
in place.
At the
end of the
inspection period,
22 months later, the licensee
had not resolved this issue.
Between
May 1993
and April 1994,
the licensee
operated with the
RUs in
service.
While the decision to valve the
RUs back into the system
was
consistent
with the guidance
provided in Generic Letter 91-18 regarding
ODs,
qualification should
have
been reestablished
in a timely manner.
The
inspector
found
11 months to be
an excessive
length of time to operate without
full qualification.
0,
1
~
-20-
The
RUs were
removed
from service
in April 1994.
The inspector
noted that
during the latter part of 1994
and early 1995,
engineering
made sporadic
progress
to resolve the qualification issue.
Although engineering
recommended
that the monitors
be abandoned,
operations
management
recommended
that they be
returned
to service.
While progress
appeared
to be
made
and
a March
1995 goal
was set,
the inspector considered
the eleven
months to be excessive
time to
resolve
the issues.
7.2.2.4
Licensee Corrective Actions
The inspector discussed
these
issues
with engineering
management.
As
a result
of the inspector's
review, the licensee initiated
a
CRDR to evaluate
the
issues
identified.
The inspector
noted that several
of the
key issues
, identified in this review had
been
addressed
to
some extent
by corrective
actions for similar issues.
~
In Nay 1994,
the licensee
implemented
an
OD proc'edure
implementing
the guidance
provided in
~
As previously noted,
the licensee
has initiated improvements
in
reportability assessments
through program enhancements
and
training.
In October
1994, the licensee
implemented training and procedure
improvements to address
weaknesses
in 10 CFR 50.59 evaluations.
In response
to several
issues
concerning timeliness
and thoroughness
of
engineering
resolution of plant deficiencies,
since August
1994,
engineering
management
has
implemented
a "Level 1" issue resolution
program,
and
has
conducted training
on the attributes of "Engineering
Excellence."
P
The inspector will review the licensee's
evaluation of this issue
and proposed
corrective actions
in a future inspection.
7.3
Current Confi uration
The inspector's
review of
EW issues
was initiated after
a weak procedure
step
was not properly followed by plant operators
(see
Section 3.2 of the report).
Theinspector
reviewed the plant configuration
and plant procedures
to
determine if there were other inconsistencies.
The inspector determined that
the control
room annunciator
response
procedure for a low surge tank level did
not indicate that automatic
makeup
from the
CT system
was disabled.
In
addition, it did not indicate
how to align the
CT system.
While the procedure
did reference
alternate
makeup
from the nonsafety-related
sources, it did not
identify the
CT system
as
a source.
t
The inspector
noted this weakness
to the system engineering
section leader.
The inspector also questioned
whether,
based
on the design basis
review
-21-
conducted after 1992,
whether it was still appropriate
to maintain the
makeup isolation valve closed.
The section leader stated that this would be
reviewed
as part of the
CRDR initiated to review the
EW configuration issues.
8
PREPARATION
FOR REFUELING
(60705)
af REFUELING ACTIVITIES
(60710)
8.1
Refuelin
Activities Unit 2
On February
12,
1995,
the inspector monitored portions of the core offload
of Unit
2 from the refueling machine inside containment.
The inspector
noted
that the refueling senior reactor operator
(SRO)
was present
on the refueling
machine.
The inspector questioned
the
SRO about the refueling machine
operations
and
TS requirements
and determined that the
SRO had strong
knowledge of refueling activities.
The inspector
observed
good
and continuous
communications
between
the refueling
and the control
room.
The inspector
noted the foreign material
exclusion controls for re.ueling operations
were
adequate.
The inspector
concluded that the licensee
demonstrated
several
strengths
in the performance of the core offload.
9
FOLLOWUP OP ERATIONS
(92901)
9. 1
Closed
Violation 50-529 93-48-02:
Failure to Follow Procedures
for
S stem
Hakeu
This violation involved the failure to return the reactor
makeup water
controller to automatic following a dilution evolution.
On November
23,
1993,
the inspector
noted that the reactor
makeup water controller was left in
manual
when the procedure directed
the operator to place the controller in
automatic following manual
operation.
The primary operator
subsequently
placed
the controller to automatic
when informed by the inspector.
As documented
in
NRC Inspection
Report 50-528/94-20;
50-529/94-20;
50-530/94-20,
the inspector
reviewed the licensee's
corrective actions
and
left this item open after finding that the licensee
had
made
a procedure
change for Unit 2 regarding the operation of the reactor
makeup water
controller,
but had not made similar changes
to procedures
in Units
1 and 3.
Subsequently,
the inspector
confirmed that the applicable
procedures
in
Units
1
and
3 had
been
changed
and closed this item.
10
FOLLOWUP - MAINTENANCE
(92902)
10. 1
Nozzle
Dam Air Hoses
The inspector
reviewed
CRDR 9-4-0107 which indicated that fittings for steam
generator
nozzle
dam air hoses
were improperly procured.
The air hoses
were
used to inflate seals
in the nozzle
dams.
Specifically, the substitute
fittings were procured
as nonquality-related
items,
and their acceptability
was not evaluated
before use.
The nozzle
dams
were classified
as quality-
related
components.
l
-22-
Prior to the Palo Verde Unit 3 midcycle outage
which began
in November
1993,
contractor
personnel
noted that hose fittings for steam generator
nozzle
dams
were difficult to use.
A member of the licensee's
refueling services
group
found
a substitute fitting and contacted
a representative
from the nozzle
dam
manufacturer
regarding
the acceptability of the substitute fitting.
The
licensee
employee
took action to procure
new fittings after being told by the
manufacturer's
representative
that the substitute fitting was acceptable.
However,
a Material Engineering
Evaluation to justify the acceptability of the
new fitting was not performed.
The fittings were procured
as nonquality-
related
even
though the nozzles
dams
were classified
as quality-related.
Procedure
" Item Procurement
Specification Requirements,"
which
applies
to quality-related
components,
indicates that
a Material Engineering
Evaluation is to be performed to justify the acceptability of substitute
items.
The nozzle
dams,
with the substitute
hose fittings, were
used during the
Unit 3 midcycle outage.
The quality classification of the fittings was
questioned
in February
1994 during
a Unit 2 midcycle outage.
The licensee
promptly initiated the
CRDR and performed
a Material Engineering
Evaluation.
The licensee
determined that the substitute fittings were acceptable
for
continued
use.
In addition,
on February
24,
1994,
the nozzle
dam manufacturer
provided written confirmation that the substitute fittings were acceptable
for
use,
The safety significance of this issue is low in that the fittings were
determined
to be acceptable,
and the nozzle
dams were designed with multiple
inflatable seals,
including
a mechanical
seal, for redundancy.
The inspector
found that the licensee
thoroughly documented this issue
and took appropriate
corrective actions,
includi'ng counseling
the individual who procured
the
fittings.
In addition,
the inspector
reviewed the licensee's Bill of
Materials database
and determined that'he
parts for the air hose
assemblies,
including the substitute fittings, were classified
as quality-related.
11
FOLLOWUP ENGINEERING/TECHNICAL SUPPORT
(92903)
Il.l
Closed
Violation 50-528 93-40-06:
Overtime Limit Exceeded
This violation occurred
when
an individual exceeded
the work hour limitations
of the TS.
The inspector
also found that licensee
audit teams
had identified
other instances
where work hour limitations had
been
exceeded,
and the
inspector
concluded that additional
management
attention to this matter
was
warranted.
On November 3,
1993,
a stop work notice
was issued to departments
with
recurring problems
in exceeding
work hour limitations.
The stop work notice
prohibited the departments
from taking exceptions
to overtime limitations and
remained
in effect until the Director, guality Assurance
(now Nuclear
Assurance),
was satisfied that corrective action plans
were developed
to
address
the problem.
Corrective Action Report 93-0179
was issued for this
matter
and
was subsequently
superseded
by
CRDR 9-4-0070.
Initial corrective
I
I
l
~
e
~
-23-
actions
included:
adding overtime limitation requirements
to general
and
continuing employee training, revising the administrative
procedure for
overtime limitations,
and improving the computer
report which identifies
overtime violations.
The inspector
found that the licensee
continues
to implement corrective
actions for this issue
and that senior licensee
management
is involved in
determining corrective actions.
NA trends
the number instances
where overtime
limits have
been
exceeded,
and the inspector
found
a declining trend in these
events.
Additionally, the majority of these
occurrences
were later determined
to be invalid due to errors in entering data from time sheets.
Currently, the
licensee
plans to develop
a
new time sheet
to ensure
consistent
reporting of
overtime
and to improve employee
awareness
of work hours.
The inspector
concluded that the licensee
has placed sufficient priority on
this matter to allow closure of this item.
The inspector will review the
licensee's
continuing corrective actions
in
a future inspection.
12
ON-SITE REVIEM OF LICENSEE EVENT REPORTS
(LERs)
(92700)
12. 1
Closed
LER 529 94-01
Revision 0:
EDG and Attendant
E ui ment
~inn erabl e
This licensee identified event occurred
when operations
personnel
did not
declare
an
EDG inoperable after the circuit breaker for the diesel
generator
building essential
exhaust
fan was
opened during maintenance activities.
The
fan was attendant
equipment to support continued diesel
operation.
At the
time of the event,
the licensee
had not implemented
a written procedure for
ODs.
Shortly after this event,
the
NRC issued
a violation to the licensee for not
declaring the low pressure
safety injection system inoperable
when
a low
pressure
safety injection
pump breaker
was
opened during
a surveillance test
(NRC Inspection
Report 50-538/94-12;
50-529/94-12;
50-530/94-12).
This
occurrence
was similar to the event with the essential
exhaust
fan in that the
licensee
disabled
equipment
and planned for manual
actions to restore
the
equipment if it was needed for operation.
Corrective actions for the
LER included issuing
a night order to operations
personnel
which indicated that the
EDG is inoperable if the breaker for the
essential
exhaust
fan is opened.
Subsequently,
the licensee
implemented its
procedure.
The inspector confirmed that copies of
procedure
were maintained
in each control
room.
The inspector
found that each unit's operability determination
logbook
included
a determination that the
EDG would be declared
inoperable if the
essential
exhaust
fan breaker
was
opened.
Also, the inspector
spoke with
control
room personnel
and found the personnel
to be familiar with the
procedure,
when credit can
be taken for manual
~ gl
I
I
i
actions to restore
equipment,
and in particular, this event.
The inspector
concluded that the licensee's
corrective actions for this event were
appropriate.
l~
ATTACHMENT 1
1
Persons
Contacted
1. 1
Arizona Public Service
Com
an
R. Fullmer, Department
Leader,
Nuclear Assurance
D. Garchow, Director, Site Engineering
B. Grabo,
Section
Leader Compliance,
Nuclear Regulatory Affairs
M. Hodge,
Group Leader,
Design Engineering
W. Ide, Director, Operations
AD Krainik, Department
Leader,
Nuclear Regulatory Affairs
J.
Levine, Vice-President,
Nuclear Production
D. Mauldin, Director, Maintenance
C.
Seaman,
Director, Nuclear Assurance
W. Simko,
Department
Leader,
Nuclear Assurance
R. Stroud,
Regulatory Consultant,
Nuclear Regulatory Affairs
J. Velotta, Director, Training
1.2
NRC Personnel
K. Johnston,
Senior
Resident
Inspector
J.
Kramer,
Resident
Inspector
A. MacOougall,
Resident
Inspector
t
1.3
Others
R. Henry, Salt River Project Site Representative
All personnel
listed
above
attended
the exit meeting
held
on March 3,
1995.
2
EXIT MEETING
An exit meeting
was conducted
on March 3,
1995.
During this meeting,
the
inspectors
summarized
the scope
and findings of the report.
The licensee
acknowledged
the inspection findings documented
in this report.
The licensee
did not identify as proprietary
any information provided to, or reviewed
by,
the inspectors.
op
I
I
0
I
gO
ATTACHMENT 2
CRDR
EW
MI
NA
TS
Condition Report/Disposition
Request
condensate
storage
tank
condensate
transfer
emergency
diesel
generator
essential
cooling water
failure data trending
high pressure
safety injection
jacket water
main feedwater
I
maintenance
instruction
motor-operated
valve
Nuclear Assurance
operations
determination
pressure
instrumentation
radiological protection
safety injection
safety injection tank
senior reactor operator
Technical Specification
V
e
~l
,f,
I
I