ML16154A568
| ML16154A568 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/16/1994 |
| From: | Harmon P, Lesser M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML16154A564 | List: |
| References | |
| 50-269-94-01, 50-269-94-1, 50-270-94-01, 50-270-94-1, 50-287-94-01, 50-287-94-1, NUDOCS 9402280114 | |
| Download: ML16154A568 (14) | |
See also: IR 05000269/1994001
Text
p
REG(,
"q
UNITED STATES
o
NUCLEAR REGULATORY COMMISSION
co
REGION II
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report Nos.:
50-269/94-01, 50-270/94-01 and 50-287/94-01
Licensee:
Duke Power Company
422 South Church Street
Charlotte, NC 28242-0001
Docket Nos.: 50-269, 50-270, 50-287, 72-4
License Nos.:
DPR-38, DPR-47, DPR-55, SNM-2503
Facility Name:
Oconee Nuclear Station
Inspection Conducted: January 2 - January 29, 1994
Inspector:
a
-'
/O
P. E. Hafmon, Senior Resident Inspector
Date Signed
W. K. Poertner, Resident Inspector
1. A. Keller, Resident Inspector
P. G
umphrey, Resident Inspector
Approved by:"
M. S. Lesser, Section Chief,
Date Signed
Reactor Projects Section 3A
SUMMARY
Scope:
This routine, resident inspection was conducted in the areas of
plant operations, surveillance testing, maintenance activities,
and engineering and technical assistance. Activities were
monitored during the Unit 3 End of Cycle 14 refueling outage.
Results:
One violation was identified, which involved inadequate procedures
controlling steam generator tube plug processes. Due to
inadequate installation, 14 tube plugs separated from their
respective tubes and were later recovered during refueling
operations (paragraph 2d).
Two Unresolved Items were identified. One URI involved improperly
sized orifice plates installed in High Pressure Injection lines
(paragraph 2c).
The second URI involved a temporary loss of power
to Unit 3 during refueling activities (paragraph 3).
Refueling activities continued on Unit 3. The resident staff
witnessed several major maintenance and modification activities
and in general considered these activities to be well planned and
executed.
9402280114 940217
PDR ADOCK 05000269
REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- B. Peele, Station Manager
S. Benesole, Regulatory Compliance Manager
D. Coyle, Systems Engineering Manager
J. Davis, Engineering Manager
- B. Dolan, Safety Assurance Manager
W. Foster, Superintendent, Mechanical Maintenance
J. Hampton, Vice President, Oconee Site
D. Hubbard, Component Engineering Manager
C. Little, Superintendent, Instrument and Electrical (I&E)
- D. Patterson, Regulatory Compliance
- S. Perry, Regulatory Compliance
- G. Rothenberger, Operations Superintendent
R. Sweigart, Work Control Superintendent
Other licensee employees contacted included technicians, operators,
mechanics, security force members, and staff engineers.
NRC Resident Inspectors
- P. Harmon
- W. Poertner
- L. Keller
- G. Humphrey
NRC Personnel
- K. Kavanagh
- Attended exit interview.
On January 21, 1994, the Chairman of the NRC conducted a routine visit
to the site, and held a press conference to discuss items of interest
with the local media. The Regional Administrator, Region II,
accompanied the Chairman on his tour and was also onsite to meet with
licensee management, and resident inspectors.
2.
Plant Operations (71707)
a.
General
The inspectors reviewed plant operations throughout the reporting
period to verify conformance with regulatory requirements,
Technical Specifications (TS), and administrative controls.
Control room logs, shift turnover records, temporary modification
log and equipment removal and restoration records were reviewed
routinely. Discussions were conducted with plant operations,
2
maintenance, chemistry, health physics, instrument & electrical
(I&E), and engineering personnel.
Activities within the control rooms were monitored on an almost
daily basis.
Inspections were conducted on day and night shifts,
during weekdays and on weekends.
Inspectors attended some shift
changes to evaluate shift turnover performance. Actions observed
were conducted as required by the licensee's Administrative
Procedures. The complement of licensed personnel on each shift
inspected met or exceeded the requirements of TS. Operators were
responsive to plant annunciator alarms and were cognizant of plant
conditions.
Plant tours were taken throughout the reporting period on a
routine basis. During the plant tours, ongoing activities,
housekeeping, security, equipment status, and radiation control
practices were observed.
b.
Plant Status
Units 1 and 2 operated at 100 percent power throughout the
inspection period. Unit 3 .remained shutdown the entire reporting
period for a scheduled refueling outage.
c.
Unit 3 High Pressure Injection Orifice
On January 15, 1994, the licensee identified that the normal and
emergency injection pressure breakdown orifice in the 3B2
injection line was a 7/8 inch (.875 inch) diameter orifice versus
a .78 inch diameter orifice as documented in the system flow
diagrams and design calculations. The increased orifice size
would result in HPI flows in excess of calculated flow rates and
could potentially result in pump runout concerns under certain
design events or reduced flow rate into the RCS assuming a failure
of an injection line.
The incorrect orifice was identified by the system engineer during
maintenance activities to correct a leaking flange connection.
The system engineer had questioned the higher flow rates seen on
the Unit 3 high pressure injection (HPI) full flow tests when
compared to the Unit 1 and 2 tests and decided to inspect the
orifice during the scheduled maintenance activity. The licensee
completed inspection of the three other orifice plates on Unit 3
and found that they were also 7/8 inch orifice plates. The
licensee was still reviewing this item at the completion of the
inspection period and had not completed an operability evaluation.
This item is identified as Unresolved Item 287/94-01-01:
Improperly Sized HPI Orifice Plates.
3
d.
Steam Generator Tube Plugs Found in Unit 3 Reactor Vessel
On January 9, 1994, while performing a post-defueling video.camera
inspection of the Unit 3 reactor vessel, the licensee discovered.
eight steam generator tube plugs on the lower grid assembly of the
reactor vessel.
During the previous operating cycle there were no
indications of higher primary-to-secondary leakage, Loose Parts
Monitor alarms, or any other indications that tube plugs had
separated from the steam generator tubesheet(s) and entered the
RCS. -Subsequent steam generator inspections revealed that there
were 14 missing tube plugs from the bottom head of the 3B Once
Through Steam Generator (OTSG).
The 14 missing tube plugs were
all part of a single batch of Inconel-690 rolled plugs installed
during the previous Unit 3 refueling outage by the Babcock &
Wilcox Nuclear Service Company (BWNS). -A review of the audio and
video records of the installation of this batch of tube plugs,
conducted by BWNS, indicated that there were 33 plugs (including
the 14 missing plugs) which may not have received the proper
torque value. Proper torque assures adequate expansion of the
plug in the steam generator tube. BWNS also reviewed all the
available tapes for Steam Generator 3A and the Unit 1 & 2 steam
generators for plugs that were installed during their last
refueling outages. Only one additional questionable tube
installation was identified. This was-located in the upper
tubesheet of Steam Generator 2B.
BWNS utilized a remote air motor control roll process to install
the tube plugs in question. The tool used in.this process was a
"ROGER" (Remotely Operated Generator Examination and Repair) roll
tool.
The ROGER roll expansion tool consisted of a torque
controlled air motor which drove the mandrel of the roll expander.
When operating properly, the amount of torque delivered to the
tube plug is achieved by adjustment of a clutch within the air
motor assembly. Upon reaching the target torque, the clutch
mechanism closes a valve within the air motor. When this valve
closes, supply air will not enter the vanes of the air motor and
the air motor turns off.
A device called the flow verification assembly was used to monitor
the air flow (by use of a rotameter) and air pressure (using a
pressure gauge) to the air motor. The rotameter was used to
ensure the air motor reached the target torque and did not stop
due to stalling out. When the target torque was reached and the
clutch mechanism turns off the air supply to the air motor, the
flow of air through the flow verification assembly stops, and the
rotameter stops rotating. If the rotameter continues to rotate,
then the air motor has stalled and the proper torque is not
assured.
The flow verification assembly also monitors the air pressure
delivered directly to the input of the air motor. This air
4
pressure is also monitored during the expansion process. The air
pressure corresponding to the proper torque is determined during
the calibration of the tool which is done prior to rolling any
plugs.
During the roll expansion process, the air pressure will
read a dynamic level, usually 5 to 10 psi below the static air
pressure. Upon reaching torque-out, the air pressure will
increase back to the static pressure and remain there for
approximately 3 seconds when the control computer shuts off the
air supply to allow retraction of the tool.
Monitoring of the flow verification assembly to assure proper
torque-out should have been done by the tooling operator during
the installation process. As stated above, proper torque-out is
assured when correct air pressure is indicated on a gage and zero
air flow is indicated by a rotameter, on the air flow verification
assembly. Video tapes of the installation of the 33 tubes in
question revealed that the pressure gauge on the air flow
verification assembly went to zero prematurely, following cut-out
of the air motor. Additionally, the characteristic sound of
proper torque-out could not be heard, and the roll times were
considerably shorter for these plug installations.
The faulty installation of tube plugs in Steam Generator 3B
occurred during a single shift by one crew. This crew installed a
total of 52 plugs in the lower head of the Steam Generator 3B
during this shift, of which 33 did not receive proper torque-out.
According to.BWNS, the individuals who performed the plug
installations had received all required training for this process
and were considered to be fully qualified to perform this task.
The controlling procedure for this activity did not provide
appropriate acceptance criteria for determining that proper torque
out was achieved, independent verification accomplished or QC
verification that proper torque-out was achieved. Furthermore,
the video tapes were not reviewed by licensee or BWNS personnel
following the activity to verify proper torque-out. Steam
generator tube plugging is an activity that is important to safety
both from the aspect of tube integrity and preventing debris
hazards to the reactor coolant system. The inspectors concluded
that the lack of independent verification and procedural guidance
to determine proper torque-out was inappropriate and resulted in
the failure to identify improperly installed steam generator tube
plugs. This matter is identified as Violation 287/94-01-02:
Inadequate Procedure for Steam Generator Tube Plugging Activities.
All 14 missing plugs were recovered. Two of these plugs were
found trapped in the lower end fittings of fuel assemblies. The
14 missing plugs and the remaining questionable plugs for Steam
Generator 3B were replaced. The questionable plug in the upper
tubesheet of Steam Generator 2B will be replaced during the next
Unit 2 refueling outage.
5
d.
Plant Tours
During a tour of the Unit 3 Reactor Building on January 7, 1994,
the inspector noted that large amounts of plastic had been used to
bag mirror insulation during the outage, 3EOC14. The insulation
had been removed from systems for work during the outage and was
inserted in plastic bags to avoid the spread of contamination.
However, the plastic material did not appear to be fire retardant.
Oconee Nuclear Site Directive 3.2.7, Control of Combustible
Materials, was reviewed to evaluate the licensee's program to
control fire loading in the reactor building. The inspector could
not find where the directive specifically addressed the fire
loading requirements in that area. The licensee confirmed that
fire loading in the reactor building was not addressed in the site
directives or procedures.
The licensee later informed the inspector that the omission of
fire loading requirements for the reactor building was being
corrected through a revision to the site directive. Based on the
licensee's corrective actions, this issue will not be opened as an
item for tracking.
f.
Condenser Circulating Water (CCW) Piping Seismic Interactions
On January 18, 1994, the licensee identified a potential piping
interaction in which the CCW pump discharge vents could be damaged
in a seismic event by the metal restraints placed around the CCW
intake piping to stabilize the lines while the piping is
unwatered. The licensee commenced to increase Keowee lake level
to above 798.1 feet to ensure gravity flow would be available to
supply suction to the Low Pressure Service Water Systems until an
evaluation could be completed to determine the seismic adequacy of
the piping configuration. In parallel with these actions the
licensee initiated exempt changes to modify the piping restraints
to increase the clearance such that adequate clearance would be
maintained during a seismic event. The licensee completed the
exempt changes to the Unit 1 and 2 CCW systems at 5:00 p.m. on
January 19, 1993. The licensee plans to modify the Unit 3 CCW
restraints prior to completion of the Unit 3 refueling outage.
This item was identified during the initiation of the problem
investigation process (PIP) to resolve drawing discrepancies in
the CCW system flow diagrams identified by the system engineer.
As a result of the PIP process, the seismic adequacy was also
questioned. The licensee had not completed the past operability
evaluation at the end of this reporting period. The inspectors
plan to review this item further after the operability evaluation
has been completed.
6
Within the areas reviewed, one violation for inadequate procedural
guidance for steam generator tube plugging activities, and one URI
concerning improperly sized HPI orifice plates were identified.
3.
Unit 3 Loss of Power to the Main Feeder Buses
At approximately 2:06 p.m., on January 27, 1994, Unit 3 experienced a
loss of electrical power to the main feeder buses (MFBs) which lasted
for approximately 21 seconds, due to PCB-59 inadvertently opening. This
resulted in a subsequent load shed of the non-vital loads from the Unit
3 MFBs. Approximately 21 seconds after the MFBs were deenergized, the
El breaker closed and energized the number 1 MFB via the startup
transformer. This action also energized the number 2 MFB since the MFBs
were connected through the ES switchgear they supplied.
The unit was in cold shutdown with all fuel offloaded to the spent fuel
pool.
The spent fuel cooling pumps were load shed during this event.
Power was restored to the spent fuel cooling pumps after approximately 4
minutes. There was no increase in spent fuel temperature. Both Keowee
units started as expected but did not tie onto the MFBs because the
buses were reenergized from the startup transformer path via the El
breaker. Units 1 and 2 were unaffected by this event.
Prior to the loss of power-event, power was being provided to the Unit 3
MFBs through the normal source breakers (NI and N2) by backfeeding
through the generator main transformer (3T) from the 525 KV switchyard
via PCB-59. A preventive maintenance activity which involved replacing
terminal strip lugs inside the number 4 Main Steam Stop Valve (MSSV)
power cabinet was ongoing. A terminal lug was inadvertently grounded
which created enough ground fault current to pick up the loss of load
relay (62GX/3), in the same 125 Vac circuit as the stop valve lugs.
Picking up the loss of load relay caused PCB-59 to open (per design).
The control room supervisor who approved the Work order (94007093- 01)
stated that he was unaware that this activity had the potential to pick
up the loss of load relay. As of the end of the inspection period,.the
adequacy of the work control process for this activity was still under
review. This matter is identified as Unresolved Item 287/94-01-03:
Maintenance Activity Results in Loss Of Power.
Within the area inspected no violations were identified.
4.
Maintenance and Surveillance Testing (62703), (61726)
a.
CCW Relay Replacement (Work Order 9400122 02)
The inspector witnessed work in progress during the changeout of
Control Relay, CCWPX2B. The changeout was a result of relay
overheating caused by a contact that did not open when required.
The work was performed per Work Order Task 9400122 02 which
implemented a special temporary instruction for the activity,
7
TI/2/1/3000/14, Temporary Instruction and Electrical Procedure To
Repair Damage Caused By Failed Coil For CCWPX2B.
The relay that was in.the CCW control circuitry affected all of
the Oconee Condenser Circulating Water Systems. To avoid a
potential loss of the normal water flows to the units, the relay
was replaced without de-energizing the circuit. This required
close co-ordination with the operations personnel and detailed
instructions for the activity.
The inspector reviewed the work activity and the required post
maintenance testing, and determined that the activity had been
carefully planned and implemented.
b.
Control Rod Drive System Test (OP/O/A/1105/09)
On January 6, 1994, the inspector witnessed operator performance
of test, OP/O/A/1105/09, Control Rod Drive System. The purpose of
the test was to verify transfer of control rod power between the
"normal" and "auxiliary" power supplies while on-line. The test
was performed in conjunction with PT/O/A/600/15, Control Rod
Movement, to verify operability of the rods with power from each
supply.
During the test, computer indication for rod Number 9 in Group 4
did not respond to the rod movement and as a result a work request
was initiated. However, rod movement was verified on the
indicating meter.
The inspector determined that the test had been properly
authorized, was performed in accordance with the procedures, and
the activities were documented as required.
c.
Calibration of LT-97 (Work Order 93057524 01)
Implementation of activities described in Work Order Task 93057524
were reviewed by the inspector on January 4, 1994.
The work
effort consisted of calibration of Differential Pressure
Transmitter LT-97. The equipment is utilized for measuring the
level in Flash Tank 3C2 and was calibrated per procedure
.IP/O/B/275/11A.
The inspectors determined the work performance and documentation
was in accordance with the applicable procedures.
d.
Turbine Driven Emergency Feedwater Test (PT/2/A/600/12)
In progress testing of the Unit 2 Turbine Driven Emergency
Feedwater Pump was observed by the inspector. The test was
performed to demonstrate operability of the pump as required by
the Technical Specifications ( TS Sections 3.4, 4.0.4, and 4.9.).
8
At the beginning of the test, an excessive amount of steam was
being emitted from the drain line. The system engineer determined
the steam was a result of condensation that had accumulated in the
steam header prior to the pump start. The steam emission stopped
as the pump continued to operate.
In addition, a cooling water leak was observed on the outboard
pump bearing. A maintenance request was made to repair the leak.
The packing was tightened which corrected the leak problem.
The inspector determined that the testing activity was performed
in accordance with the procedure and deficiencies identified were
properly dispositioned.
e.
EPSL Startup Source Voltage Sensing Circuit Test (PT/3/A/0610/01B)
This test verified proper operation of the Emergency Power
Switching Logic (EPSL) undervoltage sensing circuitry for 4160
Volt Startup Source (CT-3), proper operation of the Unit 3 Startup
Breaker Anti-Recycle (STAR) logic, and tested the contact
development for the new relays installed under TN/3/A/2886/0/0.
The test was completed satisfactorily indicating the equipment
operated as designed. The inspector noted good procedural
compliance and good communications between the technicians in the
field and the control room operators.
f.
Condenser Circulating Water Pump Discharge Pressure Test
(PT/1/A/0261/09)
The inspector reviewed PT/1/A/0261/09, Condenser Circulating Water
Pump Discharge Pressure Test, which was performed on January 26,
1994. The stated purpose of this test procedure is to demonstrate
the operability of the condenser circulating water pumps. The CCW
pumps were operated in various pump combinations and the discharge
pressures were recorded. The CCW pumps are considered operable if
recorded discharge pressures are greater than 3.48 in. HG on the
header with two CCW pumps operating. The test reviewed by the
inspector was the initial performance of the procedure and was not
a technical specification required test. The test was written as
a result of isolating the continuous vacuum priming system to the
CCW intake. The acceptance criteria was based on a design
engineering calculation that determined that 3.48 inch HG would
maintain air in solution and prevent air accumulation in the CCW
intake piping that could adversely affect operation of the
emergency condenser cooling water system.
The pressure values recorded met the acceptance criteria stated in
the test procedure. However, the acceptance criteria did not
account for operation at lower lake levels. The licensee has
established administrative limits for lake levels based on CCW
pump combinations required to maintain pressure above 3.48 inch
HG. The administrative limit for three CCW pump operation is 789
9
feet. The lake level at the time the test procedure was performed
was 797.49 feet. This item was discussed with the system engineer
and he agreed to review the acceptance criteria to determine if
normalizing the pressure indications would be appropriate.
g.
Convert 3B CST Pump to Mechanical Seals (Work Order 92014350)
Implementation of activities described in Work Order 92014350 were
reviewed by the inspector. The work effort observed consisted of
reinstalling Condensate Storage Tank Pump 3B after converting the
pump to mechanical seals as opposed to packing.
The inspectors found the work performance and documentation to be
in accordance with the applicable procedures. The inspectors did
note that the work activity was designated as non-QA. The
Condensate Storage Tank can be used as a water source to makeup to
the Upper Surge Tanks and is addressed in the Technical
specifications. The inspectors are reviewing the design
requirements of the emergency feedwater suction supplies and the
water volumes required by the TS. This item will be reviewed in
further resident inspection reports.
h.
Reactor Coolant Leakage Verification (PT/1/A/600/10)
The inspector reviewed the performance of procedure PT/1/A/600/10
conducted on January 24, 1994.
The procedure verifies reactor
coolant system (RCS) unidentified leakage was less than 1 gpm
(assuming .5 gpm evaporative losses) and calculated total combined
RCS leakage to verify operability of the Standby Shutdown Facility
Reactor Coolant Makeup Pump. The inspectors questioned the
enclosure used to calculate total combined RCS leakage in that the
enclosure did not take into account the assumed .5 gpm evaporative
loss included in the unidentified leakage calculation. The
licensee reviewed the item and revised the enclosure to account
for evaporative losses. The revision was considered an
enhancement of the procedure.
Within the areas reviewed, licensee activities were satisfactory and no
violations were identified.
5.
Engineering (71707)
Modification of Unit 3 EPSL to Automatically Close Standby Breakers in
the Event Startup Breakers Fail to Close (TN/3/A/2886/0/0):
Prior to this modification, the Emergency Power Switching Logic (EPSL)
would not automatically close the standby breakers if both startup
breakers failed to close while there was adequate voltage present at the
startup transformer. This modification installed a relay (3SBC3A,
3SBC3B) in each channel of EPSL, which is energized whenever both
startup breakers are open coincident with undervoltage on both main
feeder buses. These relays provide contacts which when closed help to
make up the circuit to energize the closing coils for the standby
breakers. The inspector reviewed the modification package and observed
cable and relay installation in the field. All activities observed were
satisfactory.
No violations or deviations were identified.
6.
Inspection of Open Items (92701) (92702)
The following open items were reviewed using licensee reports,
inspection record review, and discussions with licensee personnel, as
appropriate:
a.
(Closed) IFI 269,270,287/92-09-04, CBAST Pump Testing.
This item involved apparent inadequacies in the licensee's testing
of the Concentrated Boric Acid Storage Tank (CBAST) pumps. The
inspectors were concerned that the testing performed may not be
sufficient, in that only pump pressure and vibration were observed
and recorded. The system did not have instrumentation which would
allow flow determination. Without obtaining pump flow,
degradation and performance could not be determined. After
reviewing their program, the licensee agreed that pump flow should
be obtained, but requested relief from that requirement. The
relief request was denied, and the licensee modified the system on
all three units to provide acceptable flow instrumentation.
Testing was then performed which conformed to ASME Section XI
requirements.
b.
(Closed) URI 269,270,287/91-17-01, Density of HSM Reinforced
Concrete.
An NRC inspector previously reviewed the licensee's records for
construction of the Independent Spent Fuel Storage Facility, SNM
2503, and found that the concrete density did not meet the
requirements of the facility's TS Section 5.5. The required
density is a minimum of 145 pounds per cubic foot, which would
ensure an acceptable compressive strength. Construction records
indicated densities of various concrete building components of
approximately 144 pounds per cubic foot. The licensee submitted a
TS change which changed the requirements to a "Nominal" density
requirement having the same compressive strength. The change was
approved and implemented, clearing the inconsistency.
c.
(Closed) URI 269,270,287/92-24-01, Testing MG-6 Relay and Keowee
Overhead Path.
Following an observed failure of a breaker to close in the Keowee
Unit tie to the overhead emergency power path, the licensee found
that a permissive relay, type Westinghouse MG-6, had failed.
Further investigation determined that the relay had never been
tested, that the failure could have existed for an indeterminate
time, and the failure had resulted in Keowee Unit 2 being
inoperable when it was aligned to the overhead path. This issue is
further described in LER 269-92-014.
Following an Enforcement Conference conducted on November 8, 1993,
the NRC concluded that violations of requirements had not occurred
since the licensee had tested the relay originally and TS did not
specifically require testing of the overhead path and the
attendant MG-6 relay.
d.
(Closed) URI 269/93-03-03, Past Operability of Valve 1HP-97
During the performance of a periodic test, Check Valve 1HP-97
failed to.fully seat when pressure was applied in the reverse flow
direction.
This item was further discussed in NRC Inspection
Report 50-269,270,287/93-05 and a potential single failure concern
was identified.
The licensee performed a past operability
evaluationffor the'valve failing to fully seat and determined that
the valve had been past operable based on a mechanical inspection
of the valve internals. The NRC Office of Nuclear Reactor
Regulation reviewed the potential single failure concern and
determined that the licensee met licensing requirements for
Oconee.
7.
Review of Licensee Event Reports (92700)
The below listed Licensee Event Reports (LER) were reviewed to determine
if the information provided met NRC requirements. The determination
included: adequacy of description, compliance with Technical
Specification and regulatory requirements, corrective actions taken,
existence of potential generic problems, reporting requirements
satisfied, and the relative safety significance of each event. The
following LERs are closed:
a.
(Closed) LER 269/92-05, Equipment Failure and Defective Procedure
Result in Operation in Violation of Technical Specification. The
issues raised by this LER were dispositioned under violations 92
14-01 & 02. These violations were closed in inspection report 93
20.
b.
(Closed) LER 287/91-07, Equipment Failure Closes Pneumatic Valve
in Condensate Demi.neralizer System Causing Loss of Feedwater and
12
Reactor Trip. This event was documented in NRC Inspection Report
269,270,287/91-16.
c.
(Closed) LER 287/91-08, Excessive Reactor Coolant Leak, Reactor
Trip, and Inadvertent Protective System Actuation Result from
Management Deficiencies and Equipment Failures. This event was
documented in NRC Inspection Reports 269,270,287/91-31 and
269,270,287/91-34.
d.
(Closed) LER 269/92-14, Equipment Failure Results in the
Inoperability of Keowee Unit 2 Overhead Power Path and a Technical
Specification Violation
This issue is discussed in the closure of URI 269,270,287/92-24-01
in paragraph 6 of this inspection report.
e.
(Open) LER 269/92-17, Inadequate Seismic Support Of Vital
Instrumentation And Control Batteries Due To Unknown Cause,
Possible Installation Deficiency.
The report identified three areas associated with the 125v battery
banks where the installation of the equipment did not agree with
the applicable vendor drawings. The deficiencies involved were:
(1) a vertical support was missing on the Unit 2CB battery rail,
(2)
missing splice plates on Units 2 and 3 battery racks, and (3)
battery cells located above the butt joints on the mounting racks.
The licensee evaluated the issues and determined that the battery
rail with the missing vertical support was sufficient to meet the
seismic criteria required for an operability determination. The
determination was based on an engineering analysis that the test
data from racks with seven vertical supports had sufficient margin
such that those with six supports were acceptable.
The issue of-missing spli'ce plates where the support rails butt
together was evaluated. The results of that evaluation revealed
that the splice plates were necessary to seismically qualify the
battery racks. As a result, the missing splice plates were
installed and the installation was verified by the inspector.
The third issue involved battery cells located above the butt
joints of the mounting rails. The vendor manual was revised by the
licensee to allow batteries to be placed above the butt joints on
racks with installed seismic protection. The inspector could not
find the basis for the licensee's revision to the vendor manual
even when considering the addition of the seismic structure.
This LER will remain as an open item based on the resolution of
the third issue.
No violations or deviations were identified.
13
8.
Exit Interview
The inspection scope and findings were summarized on January 31, 1994,
with those persons indicated in paragraph 1 above. The inspectors
described the areas inspected and discussed in detail the inspection
findings. The licensee did not identify as proprietary any of the
material provided to or reviewed by the inspectors during this
inspection.
Item Number
Description/Reference Paragraph
URI 50-287/94-01-01
Improperly Sized HPI Orifice Plates
(paragraph 2.c).
VIO 50-287/94-01-02
Inadequate Procedure for Steam Generator
Tube Plugging Activities (paragraph 2.d).
URI 50-287/94-01-03
Maintenance Activity Results in Loss Of
Power (paragraph 3).