ML20244C130
| ML20244C130 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/01/1989 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20244C124 | List: |
| References | |
| 50-267-89-07, 50-267-89-7, NUDOCS 8906140134 | |
| Download: ML20244C130 (12) | |
See also: IR 05000267/1989007
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APPENDIX B
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
HRC Inspection Report: 50-267/89-07
Operating License: DPR-34
Docket: 50-267
Licensee: Public Service Company of Colorado (PSC)
P.O. Box 840
Denver, Colorado 80201-0840
Facility Name: Fort St. Vrain (FSV) Nuclear Generating Station
Inspection At:
FSV Nuclear Generating Station, Platteville, Colorado
Inspection Conducted: March 19 through April 30, 1989
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Inspectors:
R. E. Farrell, Senior Resident Inspector (SRI)
P. W. Michaud, Resident Inspector (RI)
Approved:
IN
6 - /- Ff
T. F. Westerman, Chief. Project Section B
Date
Division of Reactor Projects
Inspection Summary
inspection Conoucted March 19 through April 30. 1989 (Report 50-267/89-07)
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Areas Inspected: Routine, unannounced inspection of licensee action on
previously identified inspection findings, operational safety verification,
monthly surveillance observation, and monthly maintenance observation.
Results: Within the areas inspected, two violations were identified-
(paragraphs 4 and 6).
The licensee successfully completed an evacuation of the prestressed concrete
reactor vessel (PCRV) during this report period. This was a somewhat unique
evolution and required the use of unusual equipment and system lineups, as well
as coordination and cooperation among all licensee departments (paragraph 2).
Two instances were identified where the licensee failed to satisfy the
requirements of TS,
These examples indicate a potential breakdown in TS
compliance, peinting to weaknesses in attention to TS requirements.
Two control rods which had been replaced in February during the reserve
shutdown outage developed problems (paragraph 4). Both had recently been
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refurbished, which brings into question the maintenance and quality assurance
aspects of the refurbishment.
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An incorrect part number was specified for a replacement part on one of the
emergency diesel generators (Section 6).
This demonstrates a continuing
problem with control of maintenance activities.
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DETAILS
1.
Persons Contacted
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D. Alps, Supervisor, Security
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- F. Borst, Nuclear Training Manager
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- L. Brey, Manager, Nuclear Licensing and Resources
- M. Cappello, Central Planning and Scheduling Dept. Manager
- R. Craun, Nuclear Site Engineering Manager
- J. Eggebroten, Technical Projects Manager
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- D. Evans, Operations Manager
- M. Ferris, QA Operations Manager
- C; Fuller, Manager, Nuclear Production
J. Gramling, Supervisor, Nuclear Licensing Operations
- M. Holmes, Nuclear Licensing Manager
- F. Novachek, Nuclear' Support Manager
- H. O'Hagan. Outage Manager
- J. Re'esy, Nuclear Support Engineering Manager
- R. Sargent, Assistant to Vice President, Nuclear Operations
- L. Scott, QA Services Manager
- N. Snyder, Maintenance Dept. Manager
- P. Tomlinson, Manager, Quality Assurance
- D. Warembourg, Manager, Nuclear Engineering
The NRC inspectors also contacted other licensee and contractor
personnel during the inspection.
- Denotes those attending the exit interview conducted May 2, 1989.
2.
Plant Status
Evacuation'of the PCRV was completed on March 19, 1989.
This evacuation,
to 24.6 in. Hg., removed 182 gallons of water from the FCRV in less than
8 days. . Utilizing the normal purification system would have taken 6 weeks
or more to remove this quantity of water.
The licensee subsequently made preparations for plant startup and the
reactor was made critical on March 26, 1989.
Continued moisture removal
via the purification system, surveillance testing, and equipment repairs
were made prior to increasing power.
The turbine generator war placed on
the line April 9.
Following a Loop 1 shutdown on April 21, the exact
cause of which was not determined, the licensee shut down the reactor for
operator licensing startup tests, which occurred on April 24.
The plant was restarted on April 24 and the turbine generator placed on
the line April 27.
During weekly surveillance testing on April 27, the
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control rod drive (CRD) for Region 3 failed to scram.
The reactor was
shut down and an outage was commenced to replace the CRDs for Regions 3
and 7, which had been running at higher than normal temperatures.
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3.
Licensee Action on Previously Identified Inspection Findings
(92701)
(Closed)OpenItem(267/8722-02):
Instructions to Shift Breathable
Air Compressor Suction Not in Emergency Procedures _ - During a review of
the control room breathable air system following the October 1987 fire,
the NRC inspectors found no instructions in emergency procedures to shift
the suction of the breathable air compressors under radiological or smoke
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conditions. Procedures to perform this evolution existed, but nothing in
emergency procedures keyed personnel to those operating procedures. The
licensee revised Procedure EP-1, " Discussion of Fire," to include
Step 3.19 and Procedure EP-HZ, " Abnormal Radioactive Gas Release From
Plant," to include Step 3.2, which provide instructions to swap the
suction of the breathing air compressors. This action sufficiently
addresses the concern and this item is closed.
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(Closed)UnresolvedItem(267/8826-01): Evaluate Inconsistencies
Between Plant Drawings and the FSAR to Determine Whether Loads are
Sequenced onto 480 VAC Bus 2 - Inconsfitencies were discovered between the
FSAR, electrical drawings, and a licensee evaluation (EE-92-0008) with
regard to 480 VAC Bus 2 load sequencing. The licensee evaluated these
inconsistencies and provided a response to the NRC in a letter dated
December 15, 1988 (P-88425). Although the load sequence programs include
480 VAC Bus 2, there are no loads associated with Bus 2 which are part of
the load sequencing process. The licensee's response concluded that FSV
design documents, licensing bases, and plant drawings do not contain
inconsistencies relative to 480 VAC Bus 2 load sequencing. The NRC
inspectors reviewed the licensee's evaluation in response to this
unresolved item and consider it an acceptable basis to close this item.
4.
Operational Safety Verification (71707)
A.
General
The NRC inspectors made daily tours of the control room during normal
working hours and at least once per week during backshift hours.
Control room staffing was verified to be at the proper level for the
plant conditions at all times.
Control room operators were observed
to be attentive and aware of plant status and reasons why
annunciators were lit. The NRC inspectors observed the operators
using and adhering to approved procedures in the performance of their
duties. A sampling of these procedures by the NRC inspectors
verified current revisions and legible copies. During control room
tours, the NRC inspectors verified that the required number of
nuclear instrumentation and plant protective system channels were
opereble. The operability of emergency AC and DC electrical power,
meteorological, and fire protection systems was also verified by the
NRC inspectors. The reactor operators' and shift supervisors' logs
were reviewed daily along with the TS compliance log, clearance log,
operations deviation report (ODR) log, temporary configuration
report (TCR) log, and operations order book. Shift turnovers were
observed at least once per week by the NRC inspectors.
Information
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flow was consistently good, with the shift supervisors seliciting
comments or concerns from the reactor operators, equipment operators,
auxiliary tenders, and health physics technicians. The licensee's
station manager, operations manager, and superintendent of operations
were observed to make routine tours of the control room.
The NRC inspectors made tours of all accessible areas of the plant to
assess the overall conditions and verify the adequacy of plant
equipment, radiological controls, and security. During these tours,
particular attention'was paid to the licensee's fire protection
program, including fire extinguishers, fire fighting equipment, fire
barriers, control of flammable materials, and other fire hazards.
A walkdown of the emergency feedwater, emergency condensate, gaseous
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waste, diesel generator cooling water, and service water systems was
performed by the NRC inspectors. Valve and breaker positions were
verified where possible. When affected by a clearance, the valves or
breakers were verified to be positioned in accordance with the
clearance requirements.
Power supplies for components in these
systems were verified, but were also subject to clearances in some
cases. During these system walkdowns, the NRC inspectors verified
the operability of standby or backup equipment when components or
portions of systems were inoperable due to clearances.
B.
Radiological Controls
The NRC inspectors observed health physics technicians performing
surveys and checking air samplers and area radiation monitors.
Contamination levels and exposure rates were posted at entrances to
radiologically controlled areas and in other appropriate areas and
were verified to be up to date by the NRC inspectors. Health physics
technicians were present to provide assistance when workers were
required to enter radiologically controlled areas. The NRC
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inspectors observed workers following the instructions on radiation
work permits concerning protective clothing and dosimetry, and
observed workers using proper procedures for contamination control,
including proper removal of protective clothing and whole body
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frisking upon exiting a radiologically controlled area.
C.
Security
T5e NRC inspectors randomly verified that the number of armed
security officers required by the security plan were present. A lead
security officer was on duty to direct security activities on each
shift. The NRC inspectors verified that search equipment, including
an x-ray machine, explosive detector, and metal detector, was
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operational or a 100 percent hands-on search was conducted.
The protected area barrier was surveyed by the NRC inspectors to
ensure it was not compromised by erosion or other objects. The NRC
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inspectors observed that vital area barriers were well maintained and
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not compromised. The NRC inspectors also observed that persons
granted access to the site were badged and visitors were properly
escorted.
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D.
Review of TS LCO Compliance
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The NRC inspectors reviewed the licensee's compliance with TS Interim
LCO 3.1.8, " Reserve Shutdown System - Operation," prior to plant
.startup. Action Statement A of this LCO refers to the capability of
making an inoperable reserve shutdown (RSD) hopper operable within
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14 days following a reactor shutdown. This requirement is based on
the need to insert all RSD material, assuming all control rods fail
to insert. The shutdown margin would initially be adequate with one
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RSD unit inoperable, but under a worst case scenario, the decay of-
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protactinium (Pa) would require insertion of the RSD material from
the inoperable region within 14 days. This would be accomplished by
removing the CRD with the inoperable RSD hopper and replacing it with
a spare CRD which has an operable RSD hopper. . Experience has shown-
this can be accomplished in 2 to 3 days; hence,.the 14-day
requirement can.be met.
In addition, the licensee has a spare CRD
with an operable RSD hopper available in an equipment storage well.
During this inspection period, there were two instances in which the
licensee failed to satisfy the requirements contained in TS prior to
changing plant conditions, which caused the TS to become applicable.
The requirements of LCO 4.2.7 were violated at 10:18 p.m. on March 23
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when the PCRV was pressurized to greater than 100 psia with the
purified helium supply to the Region 27 interspace isolated.
LCO 4.2.7 specifies the PCRV shall not be pressurized to more than
100 psia unless, among other things, the interspaces between the
primary and secondary penetration closures are maintained at a
pressure greater than primary system pressure with purified helium
gas.
The purified helium supply to the Region 27 interspace was isolated
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due to Clearance No. 27419, which was not returned prior to exceeding
.100 psia. This clearance was hung on February 21, but was never
accepted for work as another clearance, No. 27851, covered these
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components. Clearance No. 27851 was returned on March 22, prior to
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exceeding 100 psia. During the licensee's review of the clearance
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log prior to exceeding 100 psia, it was noted that a clearance for
Region 27 had been returned, but the existence of a second clearance
was not discovered until 1:02 a.m. on March 24 after a new shift
crew reviewed the clearance log. Upon identifying this condition,
Clearance No. 27419 was immediately returned and the Region 27
interspace was pressurized as required.
The second instance of failing to satisfy TS requirements prior to
changing plant conditions involved a TS surveillance which was not
perfonned as required. LC0 4.6.1 specifies that the reactor shall
not be operated at power, which is defined as the linear power
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instruments indicating more than 2 percent, unless both diesel
generator sets are operable. Reactor power was increased above
2 percent on March 27 at 8:21 p.m.
It was subsequently discovered
that TS SR 5.6.1d-M, " Diesel Engine Exhaust Temperature Functional
lest," had not been performed within the previous month as required.
Reactor power was reduced below 2 percent at 1:20 a.m. on March 29.
This surveillance procedure tests the emergency diesel generators'
exhaust temperature shutdown and declutch functions. This
surveillance test had last been-performed on March 8 but was
rescheduled due to ongoing maintenance on the diesel engines.
SR 5.6.1d-M was performed satisfactorily following the power
reduction to below 2 percent on March 29, and the plant rise to power
was resumed.
The licensee was informed that these two examples of a failure to
ensure compliance with a TS prior to changing plant conditions, which
resuli.s in the TS becoming applicable, is an apparent violation of NRC
regulations (267/8907-01).
On April 18, the shim motor temperature for the Region 7 control rod
drive exceeded 250 F.
TS Interim LC0 3.1.1.B requires a temperature
of less than or equal to 250 F in order for the control rod pair to be
considered operable. Action Statement B of Interim LC0 3.1.1 allows
continued operation under these conditions provided the' shutdown
margin requirement of Interim LC0 3.1.4 is verified to be met with
the rod pair considered inoperable in its present position (fully
withdrawn). A shutdown margin calculation was performed using
SR 4.1.1. A-W/5.1.4-W, " Core Reactivity Status Check During Power
Operation." The NRC inspectors reviewed this calculation of
shutdown margin, which assumed both the maximum worth rods were stuck
and the Region 7 rod was inoperable. An acceptable shutdown margin
,was calculated and operation was allowed to continue.
A Loop 1' shutdown (ESF actuation) and reactor runback occurred'on
April 21- when a clearar.ce was being hung on the "C" Inverter which
supplies noninterruptible 120 VAC Bus 3 and Instrument Bus 3.
During
this process, power to these buses was lost. The turbine generator
immediately tripped, both Loop I helium circulators tripped, and a
Loop 1-shutdown occurred. Single channel scrams on all Channel
"C"
instruments occurred due to the loss of power to those instruments.
Power was restored to'the deenergized 120 VAC buses in approximately
30 seconds when the equipment operator hanging the clearance restored
the "C" inverter to its normal lineup. The plant responded to this
event as expected.
The licensee was unable to repeat the event during the analysis of
it. The "C" Inverter was examined and operated as designed. The
clearance (No. 25950), which was being hung, was reviewed by the
licensee and the NRC inspectors and was found to be correct. The
System Operating Procedure, SOP-92-05, Section 4.2.2, " Transferring to
Alternate Power Source at an Inverter Using the Static Transfer
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~ Switch," was also reviewed by the licensee and the NRC inspectors -
and found-to be correct. The NRC inspectors verified through
interviews of onshift personnel that the procedure was reviewed and
was in use at the time of the occurrence. The equ t. lent operator
involved walked through his actions with an offduty shift supervisor
following the event. He described no actions which would have
initiated this event. Hence, the cause has not been determined and
can be attributed to either a nonrepeatable inverter malfunction or
an operator error.
The licensee had planned to shut down on April 22 for operator
licensing examination startups, so the reactor was shut down
following the April 21 Loop 1 shutdown. The plant was restarted
after completion of the licensing startups on April 24.
During the performance of weekly 10_ inch scram tests on April 27, the
control rods for Region 3 failed to scram. When power to the brake
was deenergized, the rods dropped 0.6 inch, the rod out limit light
extinguished, and the rods stopped. Troubleshooting verified power
to the brake was being removed and that the rods could be driven in,
though drawing higher than expected current. The licensee,
therefore, initially considered the Region 3 control rods to be
" movable" but not "scrammable." This condition is not specifically
addressed in the action statements for TS Interim LC0 3.1.1 and led
to some confusion in determining action to be taken. Action
Statement A of Interim LC0 3.1.1 addresses control rods being
" inoperable due to being immovable (i.e., not capable of being fully
inserted)." Because the licensee had demonstrated the rods movable,
though not scrammable, this action statement was deemed
inappropriate.
Action Statement B of Interim LC0 3.1.1 addresses "one rod pair
trippable but inoperable due to causes other than addressed by
Action A" and provides that operation may continue if the shutdown
margin is verified with the rod pair considered inoperable in its
present position. The rods were capable of being fully inserted.
The intent of the word "trippable" was in question.
Action Statements C and D were clearly not appropriate for this
condition. Because Action Statement A was also inappropriate, and
since Action Statement B addressed "causes other than addressed by
Action A," the licensee determined Action Statement B to be
applicable and performed a shutdown margin verification as it
required. The licensee contacted the NRC Region IV office. The
licensee's inte"pretation of the actions to be taken was discussed
and it was agreed that power would not be increased, but operation
coula continue pending NRC review of the TS. The NRC inspector
learned of the situation approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> later and went to the
site to review the status of the plant. Upon reviewing the
requirements of Interim LC0 3.1.1, it became apparent that the
licensee was not truly meeting the requirements of any action
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statements of Interim LCO 3.1.1, and that, therefore, the requirements
of Interim LC0 3.0.5 were applicable.
Interim LC0 3.0.5 addresses
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the inability to meet an LCO or its action statements and requires
the plant to be in the startup mode within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in shutdown
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within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The NRC inspector discussed this with
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the licensee's station manager who then ordered an orderly plant
shutdown. The reactor was shut down with all rods verified fully
inserted at 6:54 a.m. on April 28. The initial determination occurred
at 9:30 p.m. on April 27, and, thus, the shutdown was completed
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within the time required by the applicable TS.
The licensee then entered an outage to replace the CRDs for Regions 3
and 7, which had been operating at high temperature as discussed
above. At the end of this inspection period, the control rod for
Region 3 had been removed and the outage was proceeding on schedule.
It should be noted that the CRDs from Regions 3 and 7 had recently
been refurbished and installed during the reserve shutdown outage in
February 1989.
5.
Monthly Surveillance Observation (61726)
The NRC inspectors routinely monitored the control room surveillance log
to verify that TS surveillance were complete and up to date for the
current reactor status. Additionally, the NRC inspectors monitored
surveillance activity for reactor starts and mode changes. Special
attention was paid during this inspection period to alternate cooling
method (ACM) diesel generator load tests required by TS 5.2.20.
The NRC inspectors also observed performance of the weekly fuel storage
building entry per Procedure CMG-16, Issue 3, " Entry / Exit of the Fuel
Storage Building with Fuel Present in the Facility," and security
Procedure SR-SE-16-W, Issue 10, " Fuel Storage Building Alarm System and
Key / Core Inventory." The NRC inspectors also observed the health physics
technicians performing routine source checks on the fuel storage building
radiation monitors.
During performance of SR 4.1.8.C.1/2/3-Q, " Reserve Shutdown Hopper, ACM
Disconnect, and Low Pressure Alarm Functional Test," on March 25, the
Region 34 RSD hopper failed to depressurize. The portion of this test
being performed verifies the integrity of the RSD hopper rupture disc by
pressurizing the hopper to approximately 10 psi above reactor pressure. A
pressure switch actuates to give a high hopper pressure alarm, which
indicates the hopper's integrity. A small orifice then allows the hopper
pressure to equalize with reactor pressure. This did not occur on
Region 34, as the high hopper pressure alarm did not clear. The licensee
theorized that some small piece of carbon or corrosion products might have
been clogging the orifice. The surveillance procedure was performed
again, and the hopper depressurized within the required time.
Subsequent
tests have not shown any further problems.
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On March 25, 1989, during performance of control rod testing in accordance
with SR 4.1.1.D-X
" Full Stroke Scram Test," the Region 32 control rod in
limit light did not extinguish as required. Troubleshooting by the
licenseeconfirmedtheswitch(ZSL-1211-32) was stuck shut. Each control
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rod drive has two redundant installed limit switches. A temporary change
was prepared to jumper out the defective limit switch. The NRC inspectors
reviewed TCR 89-03-05 and the associated safety evaluation. Since the
redundant limit switch (ZSL-1232-32) was demonstrated to be operational,
disabling the defective limit switch did not present an unreviewed safety
question. Compliance with TS 3.1.2, " Rod Position Indication Systems -
Operation," was verified by the NRC inspectors.
On April 27, 1989, during the performance of SR 4.1.1.B.1/2-W,
SR 4.1.2. A.3-W and SR 5.1.1.6-M, " Power / Low Power /Startup 10 Inch Scram
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Test," the control rod pair. for Region 3 failed to scram. When power to
the CRD brake was deenergized, the rod moved in approximately 0.6 inch,
the rod out limit light extinguished, and the rod motion stopped.
Troubleshooting determined the rod was movable in both the in and out
directions with the drive motor but would not scram. Actions taken in
response to this are documented in paragraph 4 of this report. The
licensee entered an outage to replace this CRD, which was in progress at
the end of this inspection period.
No violations or deviations were identified in the review of this program
area.
6.
Monthly Maintenance Observation (62703)
Replacement of cylinder heads on the "A" Diesel Generator was completed on
March 24. This was performed in accordance with Station Service
Request (SSR) 89501207 and in response to a problem with coolant flow
directors as documented in NRC Inspection Report 50-267/89-03.
9' During posteaintenance testing by the vendor on March 24, it was
determined that the replacement heads were not performing satisfactorily.
Based on the results of cylinder leakdown tests, the vendor recommended
rebuilding all 12 heads on the "A" Diesel Generator. The heads were
removed, rebuilt, and reinstalled on March 27, 1989.
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testing demonstrated satisfactory performance, and the "A" Diesel
Generator was subsequently declared operable. The licensee's
investigation of the problem disclosed the replacement heads for the "A"
Diesel Generator came from a different source than those obtained for the
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"B" Diesel Generator. The "B" Diesel Generator had previously undergone
the same repairs and performed satisfactorily during postmaintenance
tests. Additional testing of the diesel generators per the weekly
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surveillance requirements has shown satisfactory operation of the engines,
with one exception as noted below.
On April 7,1989, the "D" Engine, which is part of the "B" Diesel
Generator set, tripped on high temperature during the weekly surveillance
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test. SSR 89501818 was issued to check the calibration of the "D" Engine
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water temperature switch (TSH-92272), since it appeared the engine tripped
at a temperature lower than the setpoint of 198 12 F.
On April 12,
during troubleshooting, the water temperature switch was found out of
calibration and was replaced. The instrument and control technician
ordering the replacement part specified the wrong part number. The part
number for the oil temperature switch (which is one line above the water
temperature switch on the parts list) was erroneously selected'and
purchased. The oil temperature switch is physically the same as the water
temperature switch, and the part numbers are consecutive (1730389 and
1730390, respectively). The oil temperature switch has a range of
200-360"F, and the water temperature switch has a 130-200 F range.
The incorrect temperature switch was procured, calibrated, installed, and
tested satisfactorily on April 12, 1989. On April 19, during the weekly
surveillance test,' the "D" Engine again tripped on high temperature.
During troubleshooting efforts, it was discovered the wrong temperature
switch had been installed. The correct temperature switch was
subsequently obtained, calibrated, and installed. The licensee verified
that the correct temperature switches were installed in the other three
diesel engines. The licensee was informed that the specification and use
of an incorrect part in safety-related equipment is an apparent violation
of HRC regulations (267/8907-02).
The NRC inspectors observed the installation and calibration of a new
valve positioner on Valve PCV-4256. This valve provides reduced pressure
firewater to the diesel generators and boiler feed pump heat exchangers as
a backup source of cooling water. The existing design had a 3-15 psi
controller supplying a direct air signal to the valve actuator. This was
insufficient to keep the valve closed under all operating conditions. A
changenotice(CN-2923)andassociatedcontrolledworkprocedure
(CWP 89-0027), " Add new valve positioner with regulator to use in
combination with existing PIC-4256 for the control of PCV-4256," were
prepared by the licensee. A positioner was added which supplies 40 psi to
the valve-actuator when the controller provides a 15 psi signal. The NRC
inspectors reviewed the change notice, controlled work procedure, and
associated safety evaluation; no discrepancies were noted.
The NRC inspectors observed maintenance activities performed on
Valve HV-21257, " Emergency Condensate to ' A' Helium Circulator Water
Turbine," which had been leaking through. Because of the plant's
configuration, one side of this valve had no isolation valve, and a freeze
seal had to be utilized. The NRC inspectors observed installation and use
of the freeze seal and maintenance on the valve per SSR 89501885 and
Clearance No. 25967.
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No violations or deviations were identified in the review of this program
area.
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-12-
7.
Exit Meeting (30703)
An exit meeting was conducted on May 2, 1989, and attended by those
identified in paragraph 1.
At this neeting, the NRC inspectors reviewed
the scope and findings of the inspection.
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. _ _ . _ _ _ _ _ _ _ _ _ _ _ _