ML17311A275
| ML17311A275 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/14/1994 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17311A273 | List: |
| References | |
| 50-528-94-30, 50-529-94-30, 50-530-94-30, NUDOCS 9409220005 | |
| Download: ML17311A275 (18) | |
See also: IR 05000528/1994030
Text
APPENDIX
U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-528/94-30
50-529/94-30
50-530/94-30
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:
Wintersburg,
Inspection Conducted:
August 29 through September
2,
1994
Inspector:
C. J.
Paulk,
Reactor Inspector.
Engineering
Branch
Division of Reactor Safety
Approv
es erman,
se
,
ng)neersng
rane
Division of Reactor Safety
Ins ection
Summar
a e
Areas
Ins ected
Units
1
2
and 3:
Routine,
announced
inspection of
licensee actions taken in response to previously identified items.
Results
Units
1
2
and
3
~
The licensee
had taken actions to address all previously identified
items that were reviewed during this inspection.
Summar
of Ins ection Findin s:
~
Unresolved
Item 528;
529;
530/9307-01
was closed
(Section 2. 1).
~
Licensee
Event Report 528/93-010
was closed
(Section 2.2)..
~
Unresolved
Item 529/9312-06
was closed
(S ction 2.3).
t
~
Deviation 528;
529; 530/9317-02
was closed
(Section 2.4).
9'409220005 9409ib
ADOCK 05000528
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Violation 530/9412-04
was closed
(Section 2.5).
~
Violation 528/9412-05
was closed
(Section 2.6).
~
Inspection
Followup Item 528; 529; 530/94201-01
was closed
(Section 2.7).
~
Inspection
Followup Item 528;
529; 530/94201-02
was closed
(Section 2.8).
~
Inspection
Followup Item 528; 529; 530/94201-03
was closed
(Section 2.9).
Attachment:
~
Attachment
- Persons
Contacted
and Exit Heeting
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1
PLANT STATUS
DETAILS
During this inspection,
Unit
1 was at 98 percent
power, Unit 2 was at
100 percent
power,
and Unit 3 was at 100 percent
power until the unit tripped
on August 31 as the result of a feedwater
problem (to be documented in NRC
Report 50-528;
529: 530/94-26).
Unit 3 was returned to power on September
1.
2
FOLLOWUP
- ENGINEERING (92903)
2. 1
Closed
Unresolved
Item 50-528
529
530/9307-01:
Diverse Auxiliar
Feedwater Actuation and Antici ated Tri
Without Scram Issues
During the previous inspection,
inspectors identified three concerns
related
to the licensee's
actions related to meeting the requirements
"Requirements for Reduction of Risk from Anticipated Transients
Without Scram
(ATWS) Events for Light-Water-Cooled Nuclear
Power Plants."
These concerns
were:
1) the documentation for the software did not demonstrate
that the
software met the requirements
to be safety-related:
2) there were no
acceptance criteria available related to electromagnetic
or radio frequency
interference;
and.
3) there
was
no annunciation available
on the control board
to indicate that the system
was bypassed.
During this inspection,
the inspector
reviewed the previously identified
concerns
.
The licensee
had indicated to the
NRC that the diver se auxiliary
feedwater actuation
system
was designed
and installed
as safety-related
even
though
10 CFR 50.62 only required the system to be "reliable and independent."
The inspector
found that the licensee'had
contracted with a vendor to perform
a verification and validation of the software for the purpose of classifying
the software
as safety-related.
The inspector
concluded that the system
was
installed in accordance
with 10 CFR 50.62;
however, the licensee
had opted to
upgrade the system to safety-related.
Therefore,
there
was no violation of
NRC requirements.
The second issue,
electromagnetic
and radio frequency interference,
was
identified because
the licensee
referred to an evaluation of another
licensee's
survey without maintaining
a copy, which was considered to be
a
poor pracrice.
The inspector
noted during this inspection that evaluation of
the effects of electromagnetic
or radio frequency interference
was not
required
by industry standards
because
the
NRC has not endorsed
any such
standards.
Additionally. the
NRC issued
NUREG-0852 which contained
a
discussion of related to the performance of electromagnetic
and radio
frequency interference
surveys.
In NUREG-0852, the
NRC stated that "plant
specific testing/surveys will not be required."
The inspector
concluded that,
while such testing
may be beneficial, there
was no requi rement for the
licensee to perform the survey: therefore,
there
.i~s so violation of NRC
requirements
The lack of annunciation of a bypassed
condition of the diverse auxiliary
feedwater actuation
system
on the main control board
was identified as
an
issue
on the basis of an inspector's
interpretation of IEEE 279-1971,
"Criteria for Protection
Systems for Nuclear
Power
Generating Stations."
During this inspection,
the inspector concluded that the availability of
indication of a bypassed
condition of the system at the local panel in the
control
room met the requirements
of 10 CFR 50.62 and
While it
may have
been desirable
to have annunciation
on the main control board.
the
licensee
consciously
decided to utilize the local control
room indication
only.
In making that decision.
the licensee
accepted
the risk of the system
being bypassed
without operators
being aware of the status.
No violation of
NRC requi rements
was identified.
2.2
Closed
Licensee
Event
Re ort 528/93-010:
Ino erable Motor-0 crated
Valves in Multi le
S stems
Due to Deficiencies
Found Durin
On November 5,
1993, the licensee
determined that the capability of a number
of motor-operated
valves to perform thei r intended safety function was in
question.
The licensee
made this determination
on the basis of the results of
analysis of data obtained
from testing performed in accordance
with Generic Letter 89-10,
"Safety-Related
Motor-Operated
Valve Testing
and Surveillance.",
The inspector
found that the licensee
reached this conclusion
on the basis of
the test results.
The test results indicated that the as found torque switch
settings
were not adequate to ensure the valve would function under design
basis conditions.
This conclusion
was based
on the licensee
using
conservative
values,
derived from valve testing,
in the sizing calculations.
The inspector
reviewed
and agreed with the licensee's
conclusion that the
effect of the valves being inoperable would have been insignificant.
The
licensee
adjusted the torque switch settings in accordance with the
calculations
so that the as left settings
would ensure the valve would
function under design basis conditions.
The licensee's
motor-operated
valve program has
been inspected in accordance
with NRC Temporary Instruction 2515/109,
"Inspection Requirements
for Generic Letter 89-10. Safety-Related
Motor-Operated
Valve Testing
and Surveillance."
The inspections
were documented
in NRC Report
Nos.
50-528;
529; 530/91-25
and
94-11.
No other instances
of inoperable motor-operated
valves
have
been
identified.
2.3
Closed
Unresolved
Item 529/9312-06:
Essential
Batter
Seismic
Clearance
During
a previous inspection,
an inspector
identified
a concern related to the
installation of new batteries.
The concern
was related to a horizontal
support bracket that was less than 1.27
cm (0.5 in) from the battery cabinet.
During this inspection.
the inspector
reviewed Design
Change
Packages
2XE-PK-037,
and 3XE-PK-037,
"Replacement of Existing
, I
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Exide Model
GN Class
1E Station Batteries
EPKAF11,
EPKBF12,
EPKCF13,
and
EPKDF14 with AT8T Model KS-20472 Round Cells," Revision 0.
The inspector
noted that the change
package
provided
a maximum distance
between the
horizontal support bracket
and the battery cabinet of 45.72
cm (18 in) without
specifying
a minimum distance.
As
a result of the initial identification of
the clearance
question,
the licensee
issued Field Change
Request
E-B9378,
Amendment
NA-02.
The inspector
found that the licensee
had performed
a seismic evaluation for
the 1.27
cm distance.
The licensee
concluded that there
was
no impact on the
system
as
a result of the small distance
between the support bracket
and the
battery cabinet.
The licensee.
however,
determined that
a minimun distance
should
be specified to provide added margin.
The licensee
provided
a minimum
distance of 7.62
cm (3 in) .in Field Change
Request
E-89378.
The inspector concluded that the original design
change
package
was weak in
that
a minimum distance
was not specified.
No violation of NRC requirements
was identified since there
was no impact on the seismic design of the battery
cabinets.
2.4
Closed
Deviation 528
529
530/9317-02:
Failure to Meet Commitments
During
a previous inspection,
inspectors
identif'ied a deviation from the
licensee's
commitment to meet the guidelines of Generic Letter 89-13,
"Service
Water System
Problems Affecting Safety-Related
Equipment."
The licensee
deviated
from the commitment by not demonstrating
thermal
performance
and heat
transfer capability of both open and closed cycle heat exchangers,
and not
inspecting the service water piping on
a periodic basis
as part of the
preventive maintenance
program.
In response to the Notice of Deviation, the licensee stated that
a revised
response to the generic letter would be issued.
The inspector noted that the
revised
response
was to include
a plan to develop maintenance
tasks to
periodically inspect service water piping,
and to conduct thermal
performance
testing in accordance
with industry guidelines.
The inspector
reviewed the licensee's
revised
commitment to the generic letter
dated
September
3,
1993.
The inspector
found that the licensee
had revised
the thermal
performance testing to include only open systems
on the basis of
"high quality chemistry control," which was in accordance
with industry
guidelines.
The licensee
had inspected
the Unit 2 service water piping and
had planned
for the inspection of the Unit 1 and
3 service water piping during
refueling outages
1R4 and 3R4. respectively.
The inspector
found that the
licensee
had not completed
an evaluation of what inspections
would be required
on
a periodic basis.
The inspector also noted that the evaluation
and planned
preventive maintenance
program revisions were being tracked
on the licensee's
corrective action tracking system.
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2.5
Closed
Violation 530/9412-04:
Inade uate Post-Maintenance
Test
During
a previous inspection.
inspectors identified a failure to demonstrate
satisfactory
performance of the Unit 3, Train A emergency diesel
generator
combustion air intake valve after the completion of preventive maintenance.
The licensee
acknowledged
the violation and stated that the reason for the
violation was
a personnel
error
by maintenance
technicians
who performed work
steps
out of sequence.
A contributing factor was identified to have
been
a
fai lure to adequately
prescribe the retest
requirements
in the retest section
of the procedure.,
During this inspection,
the inspector
reviewed the licensee's
actions in
response to this violation.
The inspector
found that the licensee
had revised
procedures
and conducted training sessions
with the maintenance staff.
The
inspector
reviewed the following procedures
that the licensee
revised:
Procedure
"Maintenance Instruction Writer's Guide," Revision 10,
effective August 2.
1994:
Procedure
"Work Development
and
Control." Revision
10. effective August 17,
1994;
and Procedure
"Class
1E Diesel Generator
and Integrated
Safeguards
Surveillance Test."
Revision 6.
The inspector
concluded that the licensee
had adequately
addressed
this violation.
2.6
Closed
Violation 50-528/9412-05
Inade uate Safet
Evaluation Reviews
During- a previous inspection.
inspectors identified that
a safety evaluation
screening did not identify a change to the facility as described
in the final
safety analysis report.
As
a result, the required safety evaluation
was not
performed.
Also identified was
a safety evaluation that did not identify an
unreviewed safety question.
During this inspection,
the inspector
reviewed the revised screening
and
safety evaluation for the motor-operated
steam stop valves
and the revised
safety evaluation for the emergency
spray pond capacity.
The inspector
found
that the actions taken for the motor-operated
steam stop valves were completed
and addressed
the violation.
The licensee's
revised safety evaluation for the emergency
capacity,
completed
on July 19,
1994, did identify an unreviewed safety
question.
The unreviewed safety question
was the result of a reduction of
margin by reducing the length of time the pond is available from 27 days to
26 days.
This problem was identified as
a result of the licensee identifying
that the orifices in the return headers of the emergency
spray
pond systems
were larger than the design requirements.
In order for the licensee to
maintain
an adequate
volume of water with the increased
flow rate that would
exit with the larger orifices, the water level in the ponds
needed to be
increased.
However, with instrument accuracy taken into account.
the level
would have been
above the overflow.
The inspector
noted that the licensee
had taken actions to allow for operation
with the pond levels lower than the calculated
values.
This was accomplished
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by taking local level readings with tape measures.
thus
removing the
instrument errors.
In order to ensure the pond levels were acceptable,
the
licensee
revised
Procedure
"Routine Surveillance Daily Midnight
Logs;" Procedure
"Area 6 Operator Logs." (for Modes
1, 2,
and 3);
and.
Procedure
"Mode 5 and
6 Secondary
Area Operator
Logs."
The
inspector
was informed that the licensee
would continue this method of
monitoring unti l
a Technical Specification
change
can be processed.
The licensee
also revised
Procedure
"10 CFR 50.59 Screenings
and
Evaluations";
issued
a newsletter to all qualified screeners,
reviewers,
and
evaluators to inform them of the violation and corrective actions taken;
and,
issued Condition Report Disposition Report 94Q-056.
Additional planned
actions
included
an upgrade of the initial training and development of a
requalification training program and
an assessment
of safety evaluation
screenings.
On the basis of the completed
and the proposed actions, this item
is closed.
2.7
Closed
Ins ection Followu
Item 50-528
529
530/94201-01:
Inade uate
Batter
Si zin
Calculations
During an inspection of the licensee's
actions taken in response to
"Loss of All Alternating Current Power," inspectors
questioned
the accuracy of the licensee's
battery sizing calculations.
The inspectors
found that the licensee
used
a nonconservative
method to determine the
starting current for direct current motor -operated
valves.
During this inspection'the
inspector
reviewed the battery sizing calculation,
Calculation
13-E-PK-202.
Revision 0.
The licensee
had not completed
a revised
calculation;
however, the revised assumptions
were available.
The inspector
noted that the revised assumptions
had resulted in a conservative
method for
calculating the starting current for the di rect current motor-operated
valves
and more realistic values for modelling the current requirements
of the motor-
operated
valves.
For example.
in Calculation 13-E-PK-202,
Revision 0, the
licensee
had assumed that the motor-operated
valve would be energized with
full-load running current present for 1 minute.
In reality, the motor-
operated
valve would only be energized with a current profile that would
result in less current for approximately
4 seconds
(based
on stroke time test
results).
The licensee.
therefore,
revised the assumption
and considered
the
current required to operated
the motor-operated
valve to be eequal to locked-
rotor current for 10 seconds.
The inspector also noted that the calculated battery loads represented
approximately
60 percent of the battery capacity.
The inspector
concluded
that the licensee
had ample margin even if the original calculation
was
nonconservative.
The inspector did not identify any violation of NRC
requirements.
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2.8
Closed
Ins ection Followu
Item 50-528
529
530/94201-02:
Inade uate
Station Blackout Procedure
During the inspection of the licensee's
actioAs taken in response to
10 CFR 50.63, the inspectors
observed
a simulation of a station blackout by
licensee
operators
on the plant simulator.
The inspectors
noted that the
operating
crew did not attempt to restore the emergency diesel
generators.
The licensee
committed to revise the emergency operating procedure to ensure
that the operators
would attempt to restore the emergency diesel
generators
concurrent with the restoration of offsite power during
a station blackout
event.
The inspector
reviewed
Emergency Operating
Procedure
"Blackout,"
Revision 2, June
1,
1994:
and
Emergency Operating
Procedure
"Blackout," Revision 2. June
1.
1994.
The inspector
noted that the licensee
had
made the revisions
as committed in NRC Report 50-528;
529: 530/94-201.
The licensee
had not revised the Unit 2 procedure
because
the gas turbine
generators
had not been connected to Unit 2.
The licensee stated that the
Unit 2 procedure
would be revised to be the
same
as the other
units'rocedures.
The inspector
concluded that the licensee
had complied with the
commitment.
2.9
Closed
Ins ection Followu
Item 50-528
529
530/94201-03:
Inade uate
Fault Protection
During the inspection of the licensee's
actions taken in response to
10 CFR 50.63, the licensee
committed to review the relay settings
and revise
the corresponding
analysis of relay operations.
This commitment was
made
as
the result of the inspectors'inding
that the licensee
had not taken into
account the contribution of the gas turbine generators.
The licensee
reviewed the short circuit analysis
and found that there
was
a
small operating window to set the relays.
During this inspection.
the
inspector
reviewed Calculation AO-EC-NA-422, "Phase
and Ground Overcurrent
Relay/Breaker Selections
and Settings for
SBO System Protection Devices,"
Revision
1.
The inspector noted that the licensee
had adjusted the relays
as
a result of utilizing Curve EE-2163.
"Generator Short Circuit Decrement."
The
inspector also noted that the licensee
had revised the point at which the
fault would be sensed.
The inspector concluded that the licensee
had adequately
addressed
the
concerns identified in the previous inspection.
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ATlACHMENT
1
PERSONS
CONTACTED
1. 1
Licensee
Personnel
A. Fernandez,
Mechanical
Engineer,
Mechanical
Maintenance
Engineering
D. Garchow, Director, Engineering
A. Krainik. Manager.
Nuclear Regulatory Affairs
D. Lamontagne,
Senior Engineer,
Nuclear Regulatory Affairs
J.
Sears.
Senior
Engineer,
Nuclear Assurance
Engineering
1.2
Salt River Pro ect
R. Henry, Site Representative
1.3
NRC Personnel
K. Johnston,
Senior Resident
Inspector
The personnel
listed above attended the exit meeting.
In addition to the
personnel
listed above,
the inspector
contacted
other
personnel
during this
inspection period.
2
EXIT MEETING
An exit meeting
was conducted
on September
2,
1994.
During this meeting,
the
inspector
reviewed the scope
and findings of the report.
The licensee did not
express
a position on the inspection findings documented in this report.
The
licensee did not identify any information provided to, or reviewed by, the
inspector.
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