ML16342D466
| ML16342D466 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/22/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML16342D465 | List: |
| References | |
| 50-275-96-20, 50-323-96-20, NUDOCS 9610290211 | |
| Download: ML16342D466 (30) | |
See also: IR 05000275/1996020
Text
ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket Nos.:
50-275, 50-323
License Nos.:
Report No.:
50-275/96-20; 50-323/96-20
Licensee:
Pacific Gas and Electric Company
Facility:
Diablo Canyon Nuclear Power Plant, Units
1 and 2
Location:
7 1/2 miles NW of Avila Beach
Avila Beach, California
Dates:
August 18 through September
28, 1996
Inspectors:
M. Tschiltz, Senior Resident Inspector
S. Boynton, Resident Inspector
D. Corporandy,
Project Inspector
D. Allen, Reactor Inspector,
Region IV
Approved By:
R. Huey, Acting Chief, Project Branch E
Division of Reactor Projects
ATTACHMENT:
Partial List of Persons Contacted
List of Inspection Procedures
Used
List of Items Opened,
Closed, and Discussed
List of Acronyms
96i02902ii 96'l022
ADQCK 05000275
8
EXECUTIVE SUMMARY
Diablo Canyon Nuclear Power Plant, Units
1 and 2
NRC Inspection Report 50-275/96-20; 50-323/96-20
~Oeretinne
~
Management
demonstrated
conservative
decision making when both units were
curtailed during periods when wildland fires had the potential to jeopardize two of
the three offsite power distribution lines (Section 01.1).
Investigative and corrective actions to address
deficiencies with the controls
established for the moveable incore detector system keys during containment
entries were not timely or comprehensive
(Section R1.1).
Maintenance
~
Inadequate
work planning and procedures
were identified as significant contributors
to the deformation of spent fuel pool cooling piping during the application of a
freeze seal.
A noncited violation was identified (Section M1.1.1).
Maintenance activities associated
with charging pump relief valve weld replacement
were well coordinated
and properly performed (Section M1.1.2)
Maintenance workers did not replace tripper cams and tripper arm assembly that
had abnormal wear during overhaul of a motor-operated
damper actuator.
Failure to
replace the tripper cams resulted in the inability to place the valve in the manual
mode during subsequent
testing (Section M1.1.3).
~En ineerin
Engineering aggressively resolved equipment concerns associated
with recently
installed 4160V breaker auxiliary switches and with the chemical and volume
control system (CVCS). This was demonstrated
by the accomplishment
of several
design changes
made to improve system reliability and reduce unnecessary
operation of engineered
safety feature equipment (Sections E1.1 and E1.2).
Routine system engineer system walkdowns failed to identify and evaluate
12 CVCS system valves with evidence of packing leakage (Section E1.2).
Plant Su
ort
Procedural requirements for establishing
a fire watch were not met prior to
commencing welding in the turbine building.
A noncited violation was identified
(Section
F1
~ 1) ~
A superseded
revision of a chemistry procedure was utilized when drawing a
primary sample.
The procedure for sampling, which had been revised in April, was
used on a daily basis and was required to be verified to be the latest revision every
I 4
4
-2-
30 days.
Similar problems were noted with "issued-for-use" documents
in NRC
Inspection Report 50-275/96-06; 50-323/96-06.
A violation was identified
(Section R3.1).
There has been
a noted improvement in the general housekeeping
and radiological
work practices observed
in the Fuel Handling Building areas designated
for work on
radioactive components
(Section R8.1).
0
Re ort Details
Summar
of Plant Status
Unit 1 began this inspection period at 100 percent power.
On August 19, the unit was
curtailed to 70 percent power due to a wildland fire that threatened two of the three
500 kV transmission. lines from the plant.
The unit returned to 100 percent power on
August 20. The unit was briefly curtailed to 70 percent power on August 21, when there
was, again,
a perceived threat to the 500 kV lines from the wildland fire. Following the
curtailment, the unit returned to and remained at 100 percent power for the balance of the
inspection period.
Unit 2 began this inspection period in power ascension
at 90 percent power, following a
unit trip on August 10.
Unit 2 returned to full power on August 18.
On August 19, the
unit was curtailed to 70 percent power due to a wildland fire that threatened two of the
three 500 kV transmission
lines from the plant.
The unit returned to 100 percent power on
August 20. The unit was briefly curtailed to 70 percent power on August 21, when there
was, again,
a perceived threat to the 500 kV lines from the wildland fire. Following the
curtailment, the unit returned to and remained at 100 percent power for the balance of the
inspection period.
I. 0 erations
01
Conduct of Operations
01.1
General Comments
71707
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of
ongoing plant operations.
In general, the conduct of operations was professional
and safety-conscious.
Wildland fires that burned in areas in the vicinity of two of the three 500 kV
distribution lines created an increased potential for the loss of these lines.
Based
upon the load carrying capacity of the 500 kV line not threatened
by the fire and
the potential for grid instability, plant management
made the decision to curtail both
units to 70 percent power.
These management
actions were viewed as proactive
since they limited the potential for unplanned
plant transients
and challenges to
plant systems and operators.
02
. Operational Status of Facilities and Equipment
02.1
Effluent and Environmental Radiation Meteorolo ical Monitorin
lns ection Sco
e 71750
Radiation stack monitor recorder traces were reviewed and the operability of the
plant's meteorological indicators was audited for the period from August'2-31,
1996.
-2-
b.
Observations
and Findin
s
Primary meteorological instrumentation for the plant was out of service for the
period reviewed within the inspection scope because
of a planned replacement of
the electronic instrumentation.
The replacement was completed,
and the primary
meteorological instrumentation was returned to service on September
5, 1996.
During the period that the primary meteorological tower was unavailable, the
backup m.teorological instrumentation was opera~le in accordance
with Technical Specification (TS) 3.3.3.4, which requires one of the two channels to be operable.
Radiation monitor recorder traces were reviewed for Radiation Monitors RM-28R,
the particulate sampler; RM-14R, the noble gas activity monitor;
and RM-24R, the
iodine sampler.
Measured radiation activity levels for the monitors remained at less
than
1 percent of the radioactivity release limit during the period covered within the
scope of the inspection.
Licensee Equipment Control Guideline 39-4R7.4B,
Revision 7, which requires that either the primary or redundant
radiation monitor be
operable at all times, was satisfied for the inspected
period.
c.
Conclusions
Effluent and environmental radiation monitors and meteorological monitors met the
operability requirements
of the applicable equipment control guidelines and TS for
the period reviewed by the inspector.
Measured radiation levels for noble gas,
particulate, and iodine were significantly below alarm levels during the period,
indicating that there were no apparent uncontrolled releases of radiation during the
period.
08
Miscellaneous Operations Issues (92901)
08.1
Closed
VIO 50-323 96002-01:
failure to perform TS required channel checks of
incore thermocouples.
Licensee Procedure
STP I-1D improperly allowed the use of
the plant process computer (PPC) to perform the TS required monthly channel
checks of the incore thermocouples.
As a result, the requirements of TS 4.3.3.6
were not being met by the performance of Procedure
STP 1-1D.
The licensee researched
the use of the PPC in performing other TS required
suryeillances
and found that the PPC was also. utilized to perform channel checks of
the subcooled
margin monitor.
In response to these findings, the licensee revised
Procedure
STP I-1D to preclude the use of the PPC to perform the channel checks
of the postaccident
monitoring system instrumentation.
Subsequently,
the licensee
established
a new procedure,
STP R-27A, for conducting the monthly channel
checks of the postaccident
monitoring panel incore thermocouples
and removed that
requirement from Procedure
STP I-1D. A review of Procedure
STP R-27A found
that it adequately met the requirements of TS 4.3.3.6.
-3-
Because
Procedure
STP I-1D allowed the use of either the PPC or the local
postaccident
monitoring panel display, the licensee could not determine when
channel checks of the incore.thermocouples
had, or had not, been performed
adequately.
Consequently,
the licensee concluded that the requirements
of
TS 3.3.3.6 had not been met since initial operation of Units
1 and 2. As a result,
the licensee issued licensee event report (LER) 50-275/84-050-00.
Based upon this
review, LER 50-275/84-050-00
is closed.
08.2
Closed
VIO 50-275 95006-01:
four examples of failure to follow procedural
requirements.
The violation identified four separate
instances where opera Ions
personnel failed to follow procedures.
The licensee determined that the violations
had several different causes
including:
inattention to detail, inadequate
self-
verification, and failure to follow procedure.
One of the four examples was
subsequently
determined to not be a violation based upon additional information
that was not provided to the inspector during the initial inspection.
The inspector reviewed the licensee's corrective actions for the remaining
three examples
and determined that, in each instance, actions had been taken to
prevent recurrence of the problem,
Corrective actions included:
revision to
Operating Procedure C-7C:III, "Condensate
Polishing System Transferring Resin
Beds," issuance
of an Operations Shift Order that reiterated the requirements for
operation of sealed valves, and improvement and clarification of self-verification
criteria. The inspector concluded that the licensee's corrective actions were
appropriate.
II. Maintenance
M1
Conduct of Maintenance
M1.1
Maintenance
Observations
a.
Ins ection Sco
e 62707
The inspectors observed
all or portions of the following work activities:
~
C0146264 Unit 2 Control Room Ventilation System filter train blank flange
installation
~
PEP R-3E Replacement of Unit 2 moveable incore Detector A
b.
Observations
end Findin s
The inspectors found the work performed under these activities to be accomplished
in accordance
with procedures.
All work observed was performed with the work
package present and in active use.
Technicians were experienced
and
knowledgeable
of their assigned tasks.
The inspectors observed system engineers
-4-
monitoring job progress
and that quality control personnel were present when
required by the procedure.
When applicable, appropriate radiation control measures
were in place.
In addition, selected maintenance
observations
are discussed
below.
M1.1
~ 1
S ent Fuel Pool
Coolin
Pi in
Freeze Seal
a.
Ins ection Sco
e 62707
On August 14, after establishing
a freeze seal on a spent fuel pool cooling system
pipe, the licensee identified that the pipe had deformed in the vicinity of the freeze
seal location.
The inspector reviewed Procedure
MP M-54.3, Revision 7, "Freeze
Sealing of Piping," NMAC NP-6384-D, "Freeze Sealing (Plugging) of Piping," and
Work Order (WO) C0145900.
Actions taken to establish the freeze seal were
discussed with both the engineering personnel investigating the event and the
director of mechanical maintenance.
b.
Observations
and Findin s
Under WO C0145900, mechanical maintenance
technicians applied a freeze seal to
the SFP demineralizer/filter outlet piping, upstream of Valve SFS-1-19.
The freeze
seal was required to establish conditions needed to repack the valve. As required
by Procedure
MP M-54.3, the work planner completed Attachment 8.1 to provide
direction on the placement of the freeze seal jacket. A sketch of the pipe was
provided in accordance with Step 5.0 of the attachment
and showed the freeze seal
area centered between two pipe hangers. According to the work planner, the length
was based upon a request from the mechanical maintenance
foreman to have an
area cleaned and tested that was approximately three times the length of the jacket.
The work planner was not aware that multiple jackets were to be used for this
freeze seal at the time he completed Attachment 8.1.
To ensure the adequacy of the freeze seal, the maintenance
personnel utilized three
separate
CO, jackets side-by-side to form what they believed would be a single,
long ice plug in the pipe.
Neither engineering
nor the freeze jacket vendor were
consulted on the acceptability of this arrangement.
The as-installed configuration
was not indicated in Attachment 8.1 and was not annotated
in the WO.
In utilizing
three separate jackets, three distinct ice plugs formed.
As the ice plugs grew,
water trapped between the plugs was pressurized
and, consequently,
the yield
strength of the pipe was exceeded.
This resulted in observable
bulging of the pipe
at points between the jackets.
Both Procedure
MP M-54.3 and NMAC NP-6384-D provide guidelines for the
minimum spacing between freeze seals to protect against overpressurization
of the
water volume between the seals.
Neither document provided guidance on the use
of multiple jackets for establishment
of a single freeze seal.
For the SFP piping
-5-
application, the minimum distance between the seals should have been 3 feet.
The
actual distances between the freeze seals was less than
1 foot.
To preclude recurrence of this problem, the licensee has revised
Procedure
MP M-54.3 to explicitly prevent the use of multiple jackets for a'single
freeze seal application.
The associated
training module on freeze seals was also
revised to reflect this precaution.
The licensee assessed
the continued operability
of the deformed, piping and concluded that the deformation caused
by the freeze
seals did not significantly reduce the pipe strength and that the remaining strength,
was adequate for the application.
The affected section of pipe has been scheduled
for replacement
in October 1996.
Conclusions
The use of multiple freeze jackets resulted in the overpressurization
and deformation
of the SFP cooling piping. A violation was identified in that Procedure
MP M-54.3
did not provide adequate
guidance or controls over the attempted application.
This
licensee identified and corrected violation is being treated as a noncited violation
consistent with Section VII.8.1 of the NRC Enforcement Policy (NCV 50-
275/96020-01).
Positive Dis Iacement
Pum
Dischar
e Relief Valve
CVCS-2-RV-8116
Inlet Weld Re lacement
lns ection Sco
e 62707
The inspectors observed portions of work activities under WO C0145272,
"Replace welds on inlet piping to PD Pump Discharge Relief Valve CVCS-2-
RV-8116," and reviewed the related clearance
and history of associated
action requests
(ARs).
Observations
and Findin s
On September
11 the inspectors reviewed the WO and found the level of
detail of the instructions was adequate
for the tasks being performed.
The
instructions for the work contained the applicable requirements for prejob
briefing,.control of foreign material, cleanliness,
clearance,
permits, laydown
area, and use of tools and materials.
The maintenance
personnel were
observed to be correctly performing the instructions in the sequence
listed.
The related clearance was reviewed and the boundaries
were found to
adequately protect both equipment and personnel without unnecessarily
impacting the operability of related equipment.
The clearance tags were
hung at the specified locations in accordance with administrative
requirements.
-6-
The radiological controls for the job appeared
a'ppropriate.
Required
radiological surveys had been performed and personnel at the job-site wore
the required'protective clothing.
Potentially contaminated
areas were
correctly posted,
and appropriate actions were observed to minimize the
spread of contamination.
The tools and material at the job'site were inspected.
Welding rods were
properly tagged and controlled and they corresponded
with the work
instructions and material controls.
The test equipment was calibrated and
tagged,
and the personnel were knowledgeable
of its proper use.
The combustible material permit and welding permit were posted and
adequate
personnel
coverage was provided for the work, including
radiological protection, supervision,
and a monitor for foreign material
exclusion.
Involved personnel were knowledgeable of the necessary work
practices and the history of the equipment problems.
A comparison of Regulatory Guide 1.44, "Control of the Use of Sensitized
Stainless Steel," with the licensee's
program for control of sensitized
stainless steel indicated that the licensee's
program was consistent with the
Regulatory Guide.
The completed work documentation
was reviewed and
was noted to have been properly completed.
In addition, the training record
of the welder performing the work documented that the individual was
qualified to perform the welding.
C.
Conclusions
The observed portions of the work were noted to have been accomplished
in
accordance
with applicable procedures,
and personnel involved with the
maintenance
were knowledgeable
of procedural requirements for the work as well
as the reason for performing the design change.
M1.1.3
Overhaul of Motor 0 crated Dam er VAC-1-MOD-8
a.
Ins ection Sco
e 62707
The inspector reviewed the following work documents
and procedures:
- WO R0159930: VAC-1-MOD-8 Damper Overhaul
- MP E-53.10M, Rev 10A, "Limitorque SMB-00 and SB-00 Valve Operator
Maintenance"
- MP E-53.10A, Rev 19, "Preventive Maintenance of Limitorque Motor Operators"
-7-
b.
Observations
and Findin
s
Technical Maintenance
(TM) personnel were observed
during the installation of the
actuator worm shaft and the clutch tripper assembly for the Limitorque SMB-00
actuator that operates
Damper YAC-1-MOD-8. This portion of the work was being
repeated since, when attempting to operate the actuator following the initial
overhaul, it could not be placed in the manual mode.
Discussion with the cognizant
engineer identified that this was the first time that the actuator had been overhauled
since the plant had started operation.
Actuator Disassembl
During actuator disassembly it was noted that the tripper lever spacer had not been
installed.
Correspondence
referred to by the licensee from Limitorque indicated that
the installation of the tripper lever spacer was not required as long as proper
alignment could be obtained.
In addition, during the actuator disassembly,
an
additional spacer was noted to have been previously installed that was not specified
in the actuator diagrams.
The additional spacer had been installed between the
inner tripper cam and the bearing spacer.
After technicians consulted with
engineering,
neither the tripper lever spacer nor the additional spacer were used in
the reassembly of the actuator.
Mixed Lubricants
Upon disassembly,
technicians noted a mixture of Beacon 325 and Mobil 28 grease
in the actuator limitswitch gearbox.
Prior to reassembly,
the gearbox was cleaned
and greased
using Mobil 28 lubricant.
Discussion with the cognizant engineer
indicated that the Beacon 325 grease
had most likely been used by the valve
manufacturer since it had not been utilized by the licensee.
The mixing of grease by
addition of Mobil 28 lubricant had most likely been performed during the licensee's
prior maintenance
activities.
The combination of the two types of lubricants,
although not desired;- was evaluated
as not to have impacted valve operability since
the grease was noted to remain in a liquid state.
Although the combination of the
greases
resulted in a thinner more liquid substance,
it was considered
acceptable
since the actuator was not required to be environmentally qualified.
VAC-1-MOD-8 was one of two actuators that had not been overhauled
since
actuator maintenance
was turned over to the maintenance
department prior to the
plant commencing commercial operation.
The cognizant engineer indicated that all
other motor-operated
valve and damper actuators had been overhauled
by the
maintenance
department since that time, and the other nonoverhauled
actuator had
been satisfactorily inspected.
Based upon the information provided, there does not
appear to be a concern for the use of mixed grease
in other actuators.
-8-
Clutch Tri
er Assembl
Installation
During installation of the worm shaft, the previously installed tripper cams and
tripper fingers were inspected
and noted to be worn.
One of the tripper cams had
been slightly deformed, apparently by contact with the tripper fingers.
The tripper
fingers were also noted to be rounded, thus reducing the contact. area on the tripper
adjustment arm. The inspector concluded that a more thorough inspection of the
tripper fingers and tripper cams would have indicated the need for replacement
during initial overhaul of the actuator.
The cognizant engineer agreed that the parts
should have been replaced.
Following replacement of the tripper cams and the
tripper lever assembly, the actuator was verified to satisfactorily operate in the hand
mode.
C.
Conclusions
The initial reassembly of VAC-1-MOD-8 without replacing the worn components
is
considered to be a poor maintenance
work practice. In addition, the presence
of
mixed lubricants in the limitswitch gearbox is indicative of a prior weakness
in the
licensee's motor-operated
actuator maintenance
activities.
M1.2
Surveillance Observations
a 0
Ins ection Sco
e 61726
Selected surveillance tests required to be performed by the TS were reviewed on a
sampling basis to verify that:
(1) the surveillance tests were correctly included on
the facility schedule;
(2) a technically adequate
procedure existed for the
performance of the surveillance tests; (3) the surveillance tests had been performed
at a frequency specified in the TS; and (4) test results satisfied acceptance
criteria
or were properly dispositioned.
The inspectors observed the following surveillances:
STP l-9-L922, Revision 2, Refueling Water Storage Tank 1-1 Level Channel
LT-922 Calibration
STP P-RHR-12, Revision 4, "Routine Surveillance Test of RHR Pump 1-2"
b.
Observations
and Findin s
The inspectors found that the surveillance reviewed and/or observed were
scheduled
and performed at the required frequency.
The procedures
governing the
surveillance tests were technically adequate,
and personnel performing the
surveillance demonstrated
an adequate
level of knowledge.
The inspectors'also
noted that test results were appropriately dispositioned.
-9-
MS
Miscellaneous Maintenance Issues (92902)
M8.1
Closed
VIO 50-.275 95016-02:
Nl audio count'rate secured contrary to TS and
'rocedural requirements.
During operational tests of Nuclear Instrument
CI;annels Nl-31 and NI-32, TM personnel failed to follow procedures
and, as a
result, the audible count rate provided by these instruments was secured for
approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during a period required by TS.
Evaluating these procedural
violations, the licensee determined that the first example was caused
by personnel
error in that the technicians,
recognizing
a discrepancy
in a procedural step, did not
obtain an on-the-spot-change
(OTSC) to allow them to perform the step as desired.
The second example was most likely caused
by inadequate
self-verification.
The licensee subsequently
issued an OTSC to Procedure
STP I-4A to:
(1) allow the
audio count rate setting on the instrument to be adjusted,
as necessary,
to produce
a discernable
change
in the audio count rate, and (2) add an independent
verification requirement during the audio count rate channel restoration.
Additionally, the technicians were counseled
on the need to process
an OTSC when
work being performed is not clearly specified by the applicable procedure.
In
addition, a general tailboard was conducted with TM personnel
regarding the issues.
Prior to Refueling Outage 2R7, Procedure
STP I-4A was revised to enhance
equipment turnover requirements
contained
in the "Return to Service" section.
The
inspector verified the completion of these corrective actions and reviewed the
revised portions of Procedure
STP I-4A. The licensee's
corrective actions appeared
to be reasonable
and sufficient to preclude recurrence of the violation.
M8.2
Closed
VIO 50-275 95016-03:
failure to follow fire protection requirements for
fire doors.
During housekeeping
activities in the Emergency Diesel
Generator
(EDG) 1-1 room, personnel blocked open a fire door without informing the
fire protection specialist or the shift foreman.
The licensee determined that the root
cause of the violation was inadequate
guidance inthe governing procedure for fire
system impairment.
Specifically, Procedure
OMS.ID2 did not clearly define when a
fire door is considered
blocked or impaired. Consequently,
the personnel
performing
the housekeeping
activities did not believe that the their activities impaired the
function of the fire door.
Based upon this finding, the licensee revised Procedure
OMS.ID2 to specify when a fire door is considered impaired,
The licensee also
provided information to plant personnel
on this event, including the definition of fire
door impairment, in a March edition of the plant's Nuclear News letter. The
inspector verified the completion of the licensee's corrective actions and determined
that the actions were appropriate to preclude recurrence of the violation.
-10-
ill. En ineerin
E1
Conduct of Engineering
E1.1
4160V Vital AC S stem Review
a.
Ins ection Sco
e 71707
37551
The inspector conducted
a review of the design, maintenance,
and operation of the
on-site 4160 V vital AC power system.
This review included the following
documents:
Updated Final Safety Analysis Report (U-SAR), Chapter 8.3, Onsite Power
Systems
Design Criteria Memorandum S-63, 4160 V System
DCP H-049059, Rev. 0, 4 kV Switchgear and Cable Spreading
Room Air
Flow
DCP E-047079, Rev. 0, Setting of Secondary
Relays
DCP E-049237, Rev. 1, Tap Adjustment on the Standby Startup Transformer
DCP C-043902, Rev. 1, 4 kV Switchgear Embedment
Plate Welds
DCP M-047098, Rev. 0, 4 kV Switchgear Door Fasteners
~
TP TD-9607, Providing 125 VDC Power From SD21 to SD22 Vital Loads
Quality Evaluations: 00007408, Q0010602
~
4160 V System Surveillance Procedures
~
4160 V System Operating Procedures
The inspector also conducted
a detailed system walkdown of the 4160 V vital
switchgear and interviewed the system engineer.
This review did not specifically
address the electrical portions of the EDGs.
b, 'bservations
and Findin s
The 4160 V vital AC power system is designed to provide reliable power to various
safety-related
components.
Major loads on the 4160 V buses include emergency
core cooling system pumps, component cooling water pumps, auxiliary saltwater
pumps, auxiliary feedwater pumps, and the 480 V vital buses.
Normal power to the
buses
is fed from the unit main generator or 500 kV transmission system via the
unit auxiliary transformer.
Backup power is supplied from the offsite 230 kV
-11-
transmission system via the standby startup transformer and from the EDGs.
Each
of the three 4160 V buses has a dedicated
EDG that will automatically supply
power to the bus in the event of a loss of voltage.
Plant TS for the 4160 V system were found to be consistent with the requirements
of the Updated Final Safety Analysis Report (UFSAR). Procedures
STP M-13A,
STP M-13F, STP M-13G, STP M-13H, and STP M-75 adequately
addressed
the
requirements of TS 4.8.'l.1.1.
Operating procedures
were also found to be of
sufficient detail to provide adequate
guidance to operators in the startup, operation,
and shutdown of the 4160 V system.
Since the start of commercial operation, the licensee has identified several design
deficiencies in the 4160 V system and has proactively pursued their resolution.
Examples include the replacement of the underrated
250 MVA GE Magne-Blast
circuit breakers with 350 MVASF, breakers, improvement of the seismic withstand
capability of the 4160 V switchgear, and a modification to provide annunciation
when the breaker closing spring fails to charge.
During and following the
installation of the new 350 MVA SF~ breakers, several problems were revealed in
the interface between the new breaker and the stationary auxiliary switch.
These
problems were aggressively pursued by the licensee with corrective actions that
were both thorough and technically justified. A review of the associated
design
change
packages for these modifications found that they were technically complete
and that they adequately
addressed
any impact on the licensing basis of the
system.
The inspector walked down the 4160 V switchgear in both units, including the
ventilation lineup for room cooling.
No extraneous
materials affecting fire loading or
seismic interaction were noted in any of the switchgear rooms.
Switchgear
deficiencies were properly identified and tagged.
No deficiencies were noted that
had not already been identified.
The system engineer was very knowledgeable
on both system requirements
and
component design basis,
He has been assigned to the 4160 V system for 6 years
and is also the system engineer for the electrical portions of the EDGs.
He had
played an active rote in each of the design changes
to the system and was able to
discuss the technical details of each.
The system engineer walked down the
system on a nominal monthly basis and maintained
a quarterly system report on the
status of design changes
and resolution of deficiencies.
Conclusions
Current design and testing of the 4160 V vital,AC power system
is in conformance
with the UFSAR and plant TS. The licensee has been proactive in identifying and
correcting degraded
conditions and system design deficiencies.
Engineering,
operations,
and maintenance
staffs have demonstrated
the ability to coordinate
efforts in the implementation of design changes
and problem resolutior.
-12-
E1.2
CVCS S stem Review
a.
Ins ection Sco
e 71707
37551
The inspectors reviewed documentation
related to the CVCS, including:
ARs: A0403181,A0394714, A0394405, A0398021, A0384085,
A0384084, A0402973, A0410262, A0122862, A0314487, A0326480 and
A0393237
~
System Engineer Quarterly Reports
~
Quality Evaluations Q0011894, Q0011791, 00011639
~
Nonconformance
Report N0001955
~
UFSAR Section 6.3, Emergency Core Cooling System; Section 9.3.4,
Chemical and Volume Control System; and Table 6.2-39, Containment Piping
and Valving
~
Plant Staff Review Committee (PSRC) TS Interpretation 96-08, Revision 0
Surveillance Procedures
STP M-54, Verification of RCP Seal Injection
Flows By Resistance
Measurements,
Revision 18
Temporary Modification/Plant Jumpers 94-44, 96-14, 96-28
The inspectors walked down portions of the system in both Units
1 and 2
and observed equipment operation, valve alignments, AR tags, and overall
material condition of the equipment.
b,
Observations
and Findin s
Selected ARs on the CVCS system'were reviewed with the system engineer.
The system engineer was knowledgeable
of the status of these items and of
the equipment history for his system.
At the time of the review there was a
total, for both Units
1 and 2, of approximately 300 outstanding
ARs on the
CVCS system, some dating back to 1989.
Of the older ARs that were
reviewed,
a majority had been assigned
as low priority items.
Based on a
limited sample, the prioritization of these ARs appeared
appropriate
due to
the minor nature of the problems.
CVCS TS Re uirements
Applicable TS were reviewed.
During the review, the inspector noted that
the PSRC had approved
an interpretation of TS 3.4.6.2 e. that addressed
-13-
acceptable
CONTROLLED LEAKAGE. The TS states that the reactor coolant
system leakage shall be limited to 40 gpm CONTROLLED LEAKAGE at a
system pressure
of 2235 ~ 20 psig.
The licensee's interpretation of the TS is that its purpose
is to ensure that the
CONTROLLED LEAKAGE is less than 40 gpm under postloss of cooiant accident
conditions.
The basis for the interpretation is supported
by the TS bases, which
indicate that the limitfor controlled leakage is to ensure the safety injection flow
will be greater than that assumed
in the analysis in the event of a loss of coolant
accident.
Therefore,
in situations where the CONTROLLED LEAKAGE is calculated
to be less than 40 gpm, with charging aligned in the post accident mode, but flow
measurements
taken at 2235 psig indicate the flow rate is greater than 40 gpm, the
licensee considers the test results acceptable.
The inspector questioned the validity
of the TS interpretation in that it appeared to deviate from the surveillance as
currently written in the TS to ensure
CONTROLLED LEAKAGE at 2235 a 20 psig is
less than 40 gpm.
The most recent surveillance results for Units
1 and 2 indicated
that the CONTROLLED LEAKAGE was less than 40 gpm.
The evaluation of the
licensee's
interpretation of TS 3.4.6.2.e is being considered
as an inspection
followup item (IFI 50-275/323 96020-02).
CVCS S stem IValkdown
A walkdown of portions of the system was performed for both Units
1 and 2
and found the valve alignment to be correct, including accessible
containment isolation valves.
The operating pumps had adequate
oil levels
and cooling flows. The overall material condition of the equipment was
good, with the exception of dry boric acid indications on numerous valves.
Although most of these valves were tagged and tracked on a master AR,
12 valves were identified as having indication of dry boric acid leaks and
were neither tagged nor included in the master AR. After this concern was
raised with the licensee, the valves with evidence of boric acid leakage were
added to the master AR.
Various documents were reviewed that identified the containment, isolation
valves in the CVCS system.
The documents were consistent with the plant
drawings and with each other, with the exception of two minor editorial
errors in AD13.DC1, Attachment 7.10, "Containment tsolation Valves."
These deficiencies were identified to the licensee for correction.
The jumper log was reviewed for temporary modifications to the CVCS
system,
Three temporary pressure
instruments were noted to have been
installed to improve monitoring of the CONTROLLED LEAKAGE during
surveillance testing and to allow monitoring of the differential pressure
.
across the reactor coolant letdown filters 1-2 and 2-2 when Fi:ters 1-1 or 2-1
are isolated.
The temporary letdown filter differential pressure
gauges were
installed in April 1994 and May 1996 for Units
1 and 2, respectively.
The
-14-
use of a temporary jumper for over 2 years, in'lieu of installing a design
change,
appeared to be a protracted length of time to utilize a temporary
jumper.
However, the inspector noted that design changes
had been
initiated, and the licensee had scheduled
replacement of the temporary
gauges with permanent installations within the next month.
Centrifu
al Char
in
CC Pum
and Positive Dis lacement
Pum
Issues
The licensee has taken several positive steps to ensure that the material condition
of the CC pumps and the PD pumps is properly maintained,
as demonstrated
by the
following actions:
ao
The licensee had previously identified erosion of the restricting orifice in the
recirculation flow path for each CC pump.
Following identification,
recirculation line flow testing was performed in order to determine if the
pump recirculation flow rates were within allowable limits. Testing indicated
that CC pump recirculation flow had increased
but not to the point of
causing flow rates to be outside of allowable limits. The testing appeared to
adequately
assess the impact of the degradation
on flow rates for the
existing conditions.
The licensee has scheduled
replacement of the orifices
during the next refueling outage for each unit.
b.
In response
to industry problems described
in NRC Information
Notices 94-76: Recent Failures of Charging/Safety Injection Pump
Shafts, 80-38: Cracking in Charging Pump Cladding; and 94-63: Boric
Acid Corrosion of Charging Pump Casing Caused
By Cladding Cracks,
the licensee
has replaced one cc pump on each unit with pumps that
have stainless steel casings and internal assemblies.
C.
Following identification of an indication on a pipe weld associated
with the Unit 1 PD pump, the licensee attempted to perform an
ultrasonic examination of the Unit 2 piping.
Due to the inaccessibility
of the weld for ultrasonic examination and the potential concern for a
similar problem with the Unit 2 weld, the licensee replaced the socket
weld in question with a butt weld as a precautionary measure.
Conclusions
The licensee's initiatives to improve system reliability were noteworthy.
Significant
effort had been put forth to improve the reliability of the system and facilitate
running the PD pump to provide normal charging flow and limit operation of the CC
pumps during normal plant operation.
In addition, the system engineer was very
knowledgeable of the system and the status of outstanding
deficiencies.
One
weakness
was noted in that a number of boric acid leaks were noted that had not
be;:n identified by the licensee and entered into their tracking and evaluation AR.
-1 5-
Engineering Support of Facilities and Equipment
Review of UFSAR Commitments
A recent discovery of a licensee operating their facility in a manner contrary to the
UFSAR description highlighted the need for a special focused review that compares
plant practices, procedures,
and/or parameters to the UFSAR description.
During
the inspection period, the inspectors reviewed the applicable sections of the UFSAR
that related to the inspection areas discussed
in this report.
There were no
inconsistencies
noted between the wording of the UFSAR and the plant practices,
procedures,
and/or parameters
observed
by the inspectors.
E8
Miscellaneous Engineering Issues (92903)
E8.1
Closed
LER 50-275 94006-00:
CC pump outside of design basis due to throttling
of component cooling water (CCW) to subcomponents.
The LER was written to
report the licensee's determination that CCW flow to the CC pumps had been
reduced by throttling to the point where the CCW flow rate may not have been high
enough to adequately
cool the CC pump subcomponent
heat exchangers
in the
event of an accident to maintain postaccident
CC pump operability.
The original LER, which reported only the licensee's discovery of the problem and
immediate corrective actions, was submitted to the NRC on July 22, 1994.
Revision
1 to the LER was submitted on June 15, 1995, and addressed
the root
cause, safety significance and corrective actions for the event.
As a part of the
corrective actions, the licensee performed testing that confirmed the adequacy
of
the existing nonthrottled CCW flow to cool the CC pump subcomponents.
Although
the CCW flow to CC pump heat exchangers
remains less than that recommended
by the vendor, the licensee has consulted with the vendor and obtained concurrence
that the existing flow is sufficient to ensure the CC pump remains operable during a
design basis accident.
Final review of this issue will be performed prior to the
closure of Revision
1 of the LER.
IV. Plant Su
ort
R'I
Radiological Protection and Chemistry Controls
R1.1
Control of the Moveable Incore Detector S stem
MIDS
Ke
a.
Ins ection Sco
e 71750
In conjunction with a maintenance
observation associated
with the replacement
of
the Unit 2 MIDS Detector A, the inspec:or reviewed Procedure
RCP D-230,
Revision 9, "Radiological Control for Containment Entry" and AR A0394318 to
evaluate the radiological controls requirements for the work.
-16-
b.
Observations
and Findin s
Step 5.3.1.b.2 of Procedure
RCP D-230 requires the MIDS keys to be in the
possession
of trie'Radiation Protection
(RP) Foreman during all containment entries.
As these keys are controlled by the unit shift foreman (SFM), the RP Foreman signs
for and takes possession
of the keys prior to a containment entry.
was initiated in February 1996 to document the fact that another key controlled by
the SFM would also operate the MIDS power switch. The SFM who initiated the
AR recommended that the MIDS key lock be changed to preclude the possibility of
inadvertent operation of the MIDS drives while personnel were in the containment.
As an immediate corrective action, the SFM would subsequently
issue both keys to
the RP Foreman.
However, this action was neither proceduralized
nor documented
in the AR, and no further action was taken at that time to determine if the MIDS key
was duplicated in other applications.
On September
12, 1996, the SFM informed the RP department that there were two
more keys, controlled by the control room assistant, that would fit in the MIDS
. power switch lock. The RP General Foreman updated AR A0394318 to reflect this
discovery; however, he did not document any corrective actions taken in response.
On September
19, the inspector questioned the shift supervisor on what actions
had been taken to ensure adequate
control of the MIDS key during containment
entries.
Operations personnel
determined that the control room assistant's
keys
were initiallyturned over to the SFM and then were subsequently
removed from the
SFM's key locker and stored as spares.
Following the inspector's inquiry, the
licensee discovered that the Unit 2 MIDS power switch key was also duplicated for
use in other applications.
On September 24, in response to the number of duplicate keys identified for the
MIDS power switch, Procedure
RCP D-230 was updated to require an administrative
tagout to provide adequate
control of the MIDS during containment entries.
No
occurrences
were identified where the MIDS had been operated during personnel
entries into the containment.
C.
Conclusions
The scope of the licensee's
investigation and corrective actions in response to the
duplication of the MIDS power switch key was too narrow to adequately
bound the
- problem and ensure positive controls over the MIDS during containment entries.
This was considered
a weakness
in the licensee's corrective action program.
-17-
R3
RP&C Procedures
and Documentation
R3.1
Primar
Coolant Sam
le Procedure
P
a.
Ins ection Sco
e 71750
On September
27, the inspector observed
the drawing of a reactor coolant system
daily sample at the Unit
1 primary'ample sink.
Procedure
CAP E-1, Revision 118,
"Sampling of Primary Systems," was also reviewed.
b.
Observations
and Findin s
The "chemistry technician was knowledgeable
of the procedure
and demonstrated
proper radiological controls while working in the sample sink. A sufficient volume
of coolant was purged through the sample line to ensure
a representative
sample
was drawn.
The technician utilized an "issued-for-use"
copy of Procedure
CAP E-1, located in
the primary sample room, to draw the reactor coolant sample.
An "issued-for-use"
stamp was affixed to the cover page of the procedure,
indicating that it had been
verified to be current; however, the copy was that of Revision 11A and not 118.
The latest revision, 11B,'ad been implemented in April 1996.
A comparison
between the two revisions found only minbr administrative changes that did not
impact the intent of the procedure.
Procedure AD2.ID1, Revision 4, "Procedure
Use and Adherence," requires "issued-
for-use" procedures to be verified current.
Step 5.1.1.a states that, when a
procedure
is taken from a controlled manual and is to be used in the performance of
work, the cognizant supervisor or designated
individual shall verify that It is the
current revision immediately prior to starting work. Step 5.1.1.c states that
"procedures
in use longer than the "issued-for-use" interval shall be verified to be
the current revision..."
Both the verifier and the technician failed to identify and
update the superseded
revision of Procedure
CAP E-1 in the Unit 1 primary sample
room.
The failure to verify and update "issued-for-use" copies of controlled
procedures
was also documented
in NRC Inspection Report No. 50-275/96-06; 50-
323/96-06 with regard to the axial flux difference limits curve posted at the control
operator's station.
c.
Conclusions
The failure to verify and update the "issued-for-use" copy of Procedure
CAP E-1 in
the primary sample room was determined to be a violation of Procedure AD2.ID1
(VIO 50-275/96020-03).
-18-
R8
Miscellaneous
RPBcC Issues
R8.1
Housekee
in
in Radiolo ical Work Areas
a.
Ins ection Sco
e 71750
The inspector toured the radiologically controlled area and observed the
housekeeping
and radiological work practices in areas established to accomplish
maintenance
on contaminated
equipment.
b.
Observations
and Findin
s
The inspector observed that the radiological conditions in the 140 foot elevation of
the fuel handling building had improved from that noted in NRC Inspection
Report 50-275/96-03; 50-323/96-03.
Tools and protective clothing were noted to
be appropriately stored.
Radiological boundaries
were properly maintained in that
the areas were clearly marked and posted and there were no items laid across the
boundaries.
General cleanliness of the area had also been improved and the amount
of radioactive material that was being stored in the area had been significantly
reduced.
C.
Conclusions
The general housekeeping
and radiological work practices
in the fuel handling
building contaminated
work areas had significantly improved.
F1
Conduct of Fire Protection Activities
F1.1
Fire Watch Performance
a ~
Ins ection Sco
e 71750
On September
16, during a tour of the Unit 1 turbine building, the inspector
"'bserved maintenance
personnel performing welding on the service air supply to the
oily water separator.
The requirements of the welding and open flame permit
associated
with the work were evaluated to determine whether they were being
met.
b.
Observations
and Findin s
The Welding and Open Flame, Permit had been properly approved by a Fire
Protection Specialist, and required that a trained fire watch be stationed during the
work with a portable fire extinguisher in the work area.
Although several personnel
were in the room housing the oily water separator during the welding, a fire.watcli
was not readily identifiable.
Additionally, the required portable fire extinguisher was
outside the room, under the temporary work bench that had been set up for the job.
-19-
The portable extinguisher was, in fact, further from the work site than a permanent
fire extinguisher that was mounted on the exterior wall of the room.
The inspector
noted that the permit had not been initiated to indicate that these requirements
had
been met prior to the start of work.
Procedure
OMS.ID1, Revision 4, "Fire Loss Prevention," delineates the fire
protection requirements
during welding activities.
Section 3.3A of Attachment 7.1
to Procedure
OMS.ID'I states that the fire watch is responsible for being readily
identifiable (e.g., wearing a red vest or readily identifiable hard hat, arm band, etc.).
Section 4.3.7 states that "prior to the start of actual welding or open flame work
the worker or the fire watch shall initial the right side of the [Welding and Open
Flame Permit) after inspecting the area and confirming each of the requirements
designated
have been completed."
The failure of the maintenance
personnel to
properly designate
a fire watch and to verify that the requirements of the Welding
and Open Flame Permit had been met prior to commencing work was considered
a
violation of Procedure OMS.IDI.
co
Conclusions
The failure to properly designate
and identify a fire watch during welding activities
and the failure to verify and initial that the fire protection requirements
had been
met prior to commencing work was a violation of Procedure
OMS.ID1. This failure
constitutes
a violation of minor significance and is being treated as a noncited
violation consistent with Section IV of the NRC Enforcement Policy (NCV 50-
275/96020-04).
The placement of the portable fire extinguisher outside the room
where the welding was being performed was considered
a poor work practice.
V. Mana ement Meetin s
X1
Exit Meeting Summary
The inspectors presented
the inspection results to members of licensee management, at the
conclusion of the inspection on October 2, 1996.
The licensee acknowledged
the findings
presented.
The inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary.
No proprietary information was identified.
~
'
ATTACHMENT
PARTIAL LIST OF PERSONS CONTACTED
Licensee
R. P. Powers, Manager, Vice President
DCPP and Plant Manager
J. R. Becker, Director, Operations
D. K. Cosgrove,
Supervisor, Safety and Fire Protection
S. R. Fridley, Manager, Outage Services
W. A. Ginter, Engineer, Nuclear Steam Supply Systems
Engineering
T. L. Grebel, Director, Regulatory Services
J. A. Hays, Director, Chemistry and Environmental Services
J. R. Hinds, Director, Nuclear Quality Services
S. C. Ketelsen, Supervisor,
Nuclear Quality Services
D. B. Miklush, Manager, Engineering Services
J. E. Molden, Manager, Operations Services
M. N. Norem, Director, Mechanical Maintenance
D. A. Vosburg, Director, Nuclear Steam Supply Systems
Engineering
INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering
IP 61726: Surveillance Observations
IP 62707: Maintenance Observations
IP 71707: Plant Operations
IP 71750: Plant Support
IP 92901: Followup - Plant Operations
IP 92902: Followup - Maintenance
IP 92903: Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
~Qened
50-275/96020-01
Inadequate
work instructions and procedures for
installation of a freeze seal on SFP piping
50-275/96020-02
50-323/96020-02
50-275/96020-03
50-275/96020-04
IFI
PSRC interpretation of TS 3.4.6.2 regarding controlled
leakage
Failure to use the latest revision of CAP E-1 primary
sample procedure
Failure to follow fire watch procedures
0
-2-
Closed
50-275/96020-01
Inadequate
work instructions and procedures for
installation of a freeze seal on SFP piping
50-323/96002-01
Failure to perform required monthly channel checks of
in-core thermocouples
50-275/95006-01
50-275/9501 6-02
50-275/9501 6-03
50-275/96020-04
50-275/84050-00
Four examples of failure to follow procedure
Nuclear Instrument audio count rate secured when
required by TS
4
Fire door blocked open without authorization
Failure to follow fire watch procedures
LER
Failure to meet TS 3.3.3.6 surveillance requirements
50-275/94006-00
LER
CC pump outside of design basis due to throttling of
component cooling water to subcomponents
LIST OF ACRONYMS USED
LER
MIDS
OTSC
PSRC
SFM
TM
TS
action request
centrifugal charging
component cooling water
chemical and volume control system
licensee event report
moveable incore detector system
on the spot change
positive displacement
public document room,
plant process computer
Plant Staff Review Committee
radiation protection
shift foreman
spent fuel pool
technical maintenance
Technical Specification
Updated Final Safety Analysis Report
work order
Pq
J