ML20235V729
| ML20235V729 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/08/1987 |
| From: | Guenther S, William Orders, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20235V642 | List: |
| References | |
| 50-369-87-36, 50-370-87-36, NUDOCS 8710150295 | |
| Download: ML20235V729 (10) | |
See also: IR 05000369/1987036
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION :
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101 MARIETTA STREET,N.W.
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W Report.Nos: 50-_369/87 36,450-370/87-36
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. Licensee: Duke Power Company
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.422 South 1 Church Street-
. Charlotte,'NC- 28242
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Facility Name: McGuire' Nuclear Station' 1'and 2
- Docke'ti No( s): ! 50-369 and'50-370?
. License No(s):
NPF-9 and NPF-17:
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InspectionConductedj. August 26-September 20, 1987'
, Inspector (s):.
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'W. Oh Ers> 5enior: Revident Inspector
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5. Guenther, Resident Inspector
' Jat'e SiQned -
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. Accompanying' Personnel.:
D. Nelson
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lApproved by:
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Date Signed.
T.7 . Peebiet, S5ction* Chief
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Division of.~ Reactor Projects
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SUMMARY
Scope:
This routine unannounced: inspection involved the- areas of- operat' ions
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safety'. verification, surveillance testing, maintenance activities, followup of -
previous enforcement actions, preparations for refueling, and an emergency.
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~ preparedness exercise.
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Results:
Ir, the areas inspected, one violation was identified in the area of-
it; adequate corrective action.
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G710150295 871000 ,'
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ADOCK 05000369-
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REPORT DETAILS
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1.
~ Persons: Contacted.
Licensee Employees-
- Tl McConnell, Plant Manager
JB; Travis,l Superintendent of Operations.
.Di Rains,-Superintendent-of Maintenance-
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1B. Hamilton Superintendent of Technical Services-
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'*N.'McCraw, Compliance Engineer.
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M Sample,: Superintendent of. Integrated-Scheduling
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'*R. Banner, Compliance
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'J.:Snyder, Performance Engineer
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' D; Smith,. Performance
lOther ' licensee employees contacted- included construction craftsmen,.
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l technicians, operators, mechanics, security force members, and office
. personnel.
- Attended Exit Interview
2 ..
Exit Interview
'Thei inspection scope and findings were summarized on September 25, 1987,
-with.those persons indicated in paragraph l'above.
The licensee did not
identify 'as. proprietary any of the information reviewed by the inspectors
during the course'of their inspection.
.3.
Unresolved Items
An unresolved item (UNR) is a ' matter about which more information is
required to determine <whether it is acceptable or may involve a violation
or; deviation. One Unresolved Item concerning the deportability of an
. isolated diesel generator fuel supply valve was identified and is
delineated in paragraph 4.
. 4.
~ Plant Operations
The. inspection staff reviewed plant operations during the report period to
verify conformance with applicable regulatory requirements.
Control room
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logs, sh.i f t supervi sors' logs, shift turnover records and equipment
removal and restoration records- were routinely perused.
Interviews were
conducted with plant operations, maintenance, chemistry, health physics,
and performance personnel.
Activities within the control room were monitored during shifts and at
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shift changes.
Actions and/or activities observed were conducted as
prescribed in applicable station
administrative directives.
The
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complement of licensed personnel on each shift met or exceeded the minimum
required by. Technical Specifications.
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Plant tours taken during the reporting period included, but were not
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1.imited to, the turbine buildings, the auxilary building, Units 1 and 2
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electrical equipment rooms, Units 1 and 2 cable spreading rooms', and the
station yard zone inside the protected area.
During the plant tours, ongoing activit 6 , housekeeping, security,
equipment status and radiation control practices were observed.
a '. Unit 1 Operations
Unit i began the reporting period at full power and remained there
until August 28, when a reactor power coastdown was intitiated.
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Reactor power level and, consequently, reactor coolant system average
temperature were allowed to decrease gradually to compensate for the
-decrease in core reactivity due to fuel depletion at the end of .the
fuel cycle. On the evenin: of September 3, a controlled shutdown was
initiated from approximately 85 percent power, and the unit was taken
off line early on the morning of September 4, to begin its fourth
refueling outage.
The reactor coolant system was cooled down to Mode 4 (hot shutdown)
on September 4 and to Mode 5 (cold shutdown) on September 5.
Mode 6
.(refueling) was entered on September 10, and core alterations were
commenced on September 20.
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Each unit is equipped with two emergency diesel generators (EDGs) to
supply standby power for safe unit shutdown in the event of a loss of
of f site oower.
Each diesel generator has an independent fuel oil
storage tank and supply system.
When required for maintenance
access, the 50,000 gallon underground storage tank can be isolated by
shutting one of two valves in the supply line connecting the storage
tank to the remainder of the fuel oil system. One valve is interior
to its corresponding diesel room while the other is outside ' the
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diesel room in the underground storage tank area and is in close
proximity to the exterior valve for the other diesel of that unit.
On September 11, at approximately 6:10 p.m. , the licensee discovered
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that the exterior fuel oil supply valve (1FD-74) to diesel generator
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1B was mistakenly shut instead of the exterior fuel oil supply valve
(1FD-67) to diesel generator 1A.
This condition, causing diesel 18
to 'be inoperable, had existed since approximately 11:50 p.m.
on
September 8, during which time diesel 18 had been considered the only
operable diesel for Unit 1 as diesel 1A was undergoing major outage
maintenance.
Unit I was in Modes 5 and '6 during this time.
The
valve alignment was corrected at the time of discovery by the
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licensee.
The cause of the mistaken valve alignment was incorrect
valve identification in that the label plates for the two valves
were interchanged. The licensee determined that the label plates were
initially installed in this configuration approximately six months
prior to the occurrence when Operating Procedure OP/1/A/6350/02
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'(Diesel . Generator) was used to verify the valves' location.
This
procedure incorrectly identified the valves, interchanging their
location.
The licensee corrected the ' label _ plates at the time of
' discovery and promptly issued a change to correct the operating
. hereas the interior supply valves are normally : shut for
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procedure.
isolation instead of the exterior valves, it is considered unlikely
that the. exterior valves had previously been inappropriately operated
.during the time'the valves were misidentified.
Having both diesel generators inoperable in Modes 5 and 6 constitutes
a violation 'of Technical Specifications . 3.1.2.1,
3.1.2.3,
and
3.8;1.2.
Furthermore, the procedural inadequacies delineated above,
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concerning operating procedure OP/1/A/6350/02 constitute a violation
of Technical Specification 6.8.1.
It should also be noted that since
the c unit had not reached the point of moving fuel, the action
statements of the above specifications were unknowingly met.
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Inasmuch as the circumstances of the above event satisfy the
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requisites del.ineated in 10 CFR 2 Appendix C concerning licensee
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identified violations, no Notice of Violation will be issued. (LIV
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50-369/87-36-01)
In a related matter, the ' subject of the deportability of the above
described incident will be carried as an Unresolved Item pending
completion of the licensee's evaluation.
Currently, the licensee
does not consider the event to be reportable because, unbeknownst to
them at the time, the . action statements of the applicable
specifications were met. (UNR 50-369/87-36-02)
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The unit was in Mode 6, with the reactor coolant system drained for
"mid-loop" operations and the reactor pressure vessel head removed, .
on the morning of September 16, when a loss of offsite power
occurred.
One of the two independent circuits between the offsite
transmission network and the onsite essential auxiliary power system
had been removed from service for outage maintenance as permitted by
the plant's Technical Specifications in the existing mode.
The
second offsite power source was lost at 10:15 a.m. as a result of
nuclear station modification testing being performed by Instrument
and Electrical (IAE) personnel .
All systems appeared to respond
normally to the loss of offsite power. The operable diesel generator
started and loaded as required, and the operators promptly restarted
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the residual heat removal (NU) pump on emergency power
Reactor
coolant system temperature increased about six degrees during the
four minutes in which no ND flow existed. Offsite power was restored
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at 10:25 a.m.
At 10:30 a.m. the licensee declared a Notification of
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Unusual Event (NOUE) as required by the emergency plan; the NOUE was
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terminated at 1:49 p.m.
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The licensee has initiated a Problem Investigation Report (PIR Serial
No. 0-M87-0208) to follow-up on this event and determine appropriate
corrective actions to prevent recurrence. The inspectors will review
the licensee's PIR and track this event as an Inspector Follow-up
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Item (IFI 50-369, 370/87-36-03).
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LUnit 21 operated atLfull .. power from; the beginning of the reporting
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- period . unti1 the'. morning ofy September 1,1987, when ..the 2A main
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afeedwaterl pump' tripped. causing a runback to: approximate 1y'50. percent
power.; The feed pump trip was caused by moisture entering the local-
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'feedwater pump turbine. control cabinet / circuitry. JanitorialJservice
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jpersonnel =were' cleaning: the turbine operating deck one elevation
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' above' the main. feedwater pump and had l failed to adequately -seal : the
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(equipment . hatch used to allow. overhead crane: access to..the :feedwater
pump . A small amount of' cleaning fluid leak'ed 'under the : tape' which .
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had been applied to seal the crevice around the hatch and dripped on
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Lthe turbine control cabinet. causing an electrical malfunction. All
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systems responded normally during- the runback and the unit was'
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. returned to full power ~.later that day, after restoring feed pump
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. operability.
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iThe. Unit : continued to operate at 100 percent power until 10:35 a.m.
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' on September 6, when- it tripped on high. pressurizer pressure. The
sequence;of events began when the "A" instrument air (VI) compressor-
was restarted after undergoing maintenance. .Tne compressor motor.had
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. a shorted;1ead which resulted in a trip of load center SMXT, which
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-feeds; motor - control centers MKB, KXB, . and KRB.
Loss of these
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electrical supplies ultimately resulted in;the trip of a main turbi.ne-
control ~ interface. relay, GBX, which was perceived: by' the turbine
control logic as a total loss of main ' generator load and caused the
turbine' governor valves to close.to prevent a turbine overspeed. The,
lossiof ca' secondary heat sink caused primary system temperature and
.' pressure.to increase to the high pressurizer pressure trip setpoint
of. 2385 psig.
- The' -lost. electrical supplies _ also prevented
pressurizer power . operated relief valves, condenser steam dumps
and atmospheric. dumps from functioning, thereby
increasing the
magnitude o_f the' primary pressure transient. Pressure peaked at 2412
psig, .so the' primary system code safety valves, which have lift
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setpoints of 2485 psig, were not. challenged. The. event was further
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complicated by a loss of.. throttling capability for the auxiliary
feedwater system and by the moisture. separator / reheater extraction
' steam valves remaining open after the trip.
These complications,
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which were also attributable to the loss of power, caused difficulty
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in controlling the primary coolant system cooldown rate and resulted
. i n 'a decrease inl average coolant _ temperature to 540 degroes-
Fahrenheit.
Power to the lost load centers was restored at about
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10:44 a.mi, facilitating a normal plant recovery from the reactor
. trip.
The circumst'ances surrounding this event, and in particular
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the coordination of electrical breakers, will be carried as an
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Inspector Follow-up Item.
(IF1 50-370/87-36-04)
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The unit . was restarted on September 7 and operated at. full power
until September 16, when a loss of instrument air (VI) pressure
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caused the main feedwater regulating valves to close and the reactor
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to trip on low steam' generator water level.
The plant responded
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normally during the trip and was placed back on the line early on the
mor.ning of September 17; it completed the reporting period at full
power,
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' The September'16,. loss of. instrument air resulted from the' loss of
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offsite power event discussed in the Unit 1. Plantioperations section-
'of , this. report,
Both McGuire units' share an instrument air' system-
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. hich is comprised of. .four compressors, . associated' air dryers. and -
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receivers. : Three of the four. compressors were aligned to receive
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their electrical power 'from Unit 1 on1 the morning of. September '16,
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thus when offsite power was lost, the single-surviving' compressor was
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zi ncapable of supplying system demand.
Approximately six minutes
after the' loss of offsite power on' Unit 1, VI pressure decreased to:a
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point ~ at which the Unit 2 air-operated feedwater regulating valves
began to elose .and steam ; generator water level began. _to . drop. The
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operators were unable-to restart.the VI compressors on- Unit 2 power
soon Anough .to halt the loss of VI pressure. As noted-in the~ Unit 1
. operations' discussion, the licensee -has inititated a PIR r to
investigate this event and the inspectors will followup on the
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licensee's corrective actions.
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During a: tour .of the Unit 2 auxiliary feedwater (CA) pump rooms on
September 1,.1987, the inspectors noted that 2CA-87, a 2A motor
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driven pump manual discharge, isolation' valve, had a lock and chain
attached to the-valve yoke.that did not in any way capture the valve
handwheelfto prevent it from being operated. The inspectors verified
that .the valve was open and proceeded to the control room to
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determine from - the CA': procedure (0P/2/A/6250/02) valve checklist
whether the valve was supposed to be locked in the open position and
to . report the apparentL discrepancy to the Operations staff,
The
procedure confirmed that 2CA-87 was supposed to be locked in the open
position. ~ An Operator was promptly dispatched to <orrect the
deficiency.
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Any deviation from normal system alignment is..normally controlled by.
the Tagout/ Removal and Restoration (R&R) procedure, OMP 2-17.
Accordingly, the R&R index was reviewed to determine whether an R&R
had recently been issued for the CA system.
It revealed that R&R
number 27-286 had been issued .on the 2A motor driven CA pump on
August 27 and that sequence number 3 of the tagout required that
2CA-87 be closed. When the system was restored later that day, the
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return position block for 2CA-87 indicated that the valve should be
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returned to the open position, but it failed to indicate that the
valve should be locked, as' required by 0P/2/A/6250/02 and OMP 2-17.
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NRC Inspection Report Nos. 50-369,370/87-26 documents a reactive
inspection conducted to review the' circumstances surrounding a Unit 1
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diesel generator inoperability which occurred between July 26 and 30,
1987.
One factor involved in that incident was the apparent
inadequacy of an R&R processed to remove an RN pump from service for
preventive maintenance. The equipment tagged block of the R&R failed
to adequately specify how the diesel generator cooled oy that RN pump
was to be prevented from starting (for eq-ui pment orotection
purposes).
This incident is being considered for escalated
enforcement action,-and the Notice of Violation has nct yet been
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issued by the NRC.
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Since the licensee has not yet been cited nor had a chance to respond
toithis earlier-apparent violation of the R&R procedure, no Notice of
Violation _will be proposed for this similar finding.
c.
Unit 2 Fuel Element Defect
Unit.2 completed 'its last' refueling outage on July 6,
1987, and
opercted with. normal
fuel
performance until August 26, when
indications of a possible fuel element defect were detected.
The
reactor cociant system dose equivalent iodine-131 concentration had
been running at normal levels (i.e., less than 0.05 microcuries/ gram)
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but showed a suoden increase to approximately 0.31 microcuries/ gram.
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The sample . frequency has been increased to help trend the variations
in iodine concentration, particularly after operational transients.
The highest peak recorded to date has been 0.93. microcuries/ gram,
which remains below the Technical Specification limit of 1.0
microcurie / gram.
Personnel at McGuire and the Duke Power Company General Office have
been tracking the data closely and have evalcated the consequences of
operating for an extended period of time with a significant fuel
element defect. The licensee has compared the limited data available
for Unit 2 at this time with the data from fuel element defects at
other Duke facilities and other licensees.
Historically, fuel
element defects have been induced by two mechanisms, either grid
fretting or debris. The data currently available appears to fit the
coolant activity trend found in debris induced fuel failures.
The
licensee is continuing to track changes in coolant activity, with
samples being taken on twelve hour intervals,
The inspectors will
continue to monitor licensee activities in this area.
5.
Surveillance Testing
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Selected surveillance tests were analyzed and/or witnessed by the
inspector to ascertain procedural and performance adequacy and conformance
with applicable Technical Specifications.
Selected tests were witnessed to ascartain thst current written approved
procedures were available and in use, that test equipment in use was
calibrated, that test prerequisites were met, that system restoration was
completed and test results were adequate.
Detailed below are selreted tests which were either reviewed or witnessed:
PT/1/A/4208/03
Containment Spray Heat Exchanger Integrity Test
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IP/0/A/3001/01
Main Steam Flow Calibration
NRC Inspection Report Nos. 50-369,370/87-05 documented a violation of
Technical Specification 6.8.1 involving a deficiency in the Slave Relay
Test procedures, PT/1 and 2/A/4200/28.
The procedures failed to
adequately specify the nuclear service water (RN) and control area
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, ventilation and. chilled water (VC/YC) system alignments necessary to?
satisfactorily.. accomplish the periodic . testing. of safety injection slave
relays and prevent the unnecessary protective' trip of the VC chillers due
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to'a loss of' cooling water flow.
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In.;its -response' to this violation, the licensee stated that the
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appropriate sections;of the PT had been changed to' ensure that this event'
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could notl recur land,. additionally, that the procedures were reviewed for
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!similar shortcomings and none'were noted.
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'On' July 21,11987,- both NcGuire units were operating.at full power and the
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' "A": train .of. VC/.YC wasL supplying control room ventilation. The normal RN
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essential header discharge flow path ' to the main .. condenser circulating
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Jwater system discharge was isolated (i.e., valve ORN-147A was closed and
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0RN-148A-was open) and. valve 0RN-149A was open sending the RN discharge to
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the standby nuclear service water pond (SNSWP). The Performance group was
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making preparations to . conduct "A"
train safety injection slave relay :
testing in accordance with section 12.25 of PT/2/A4200/28, which required
valve ORN-147A to'be in the open position but failed to'specify a position
for ORN-148A. At Performance's request the Operations staff realigned the
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RN valves to comply with the PT and .sti'll maintain the existing RN.
flowpath to the SNSWP' (i.e. , valve ORN-147A was opened and ORN-148A was
! closed).
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. When the train
"A" safety injection slave relays (K610 and K641) were
energized,r valve ORN-149A closed as required, thereby isolating the only
avaliable discharge flowpath for the essential RN header. This. caused the
operating "A".: train.VC/YC chiller to trip on loss of cooling water flow as
the "B" train had done on February 4, 1987.
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A review of the test procedure indicated that ste'p 12.25.2 of the PT had
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directed the technician to ensure that RN pump 2B was supplying any
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requ. ired."B" train essential loads. RN pump 2B was . running at the time,
however, since the RN supply _ header was cross connected, it was supplying
flow to the, "A"
VC/YC chiller, an essential
"A" trai n load.
A very
similar statement in- section 12.26 of the same procedure was responsible
for the incident on' February 4.
It had also directed the technician to
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ensure that RN pump 2A was off and that all required loads were being fed
from train'"B".
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Criterion XVI,
requires
that measures
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.' established to assure that conditions adverse to quality, such as
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deficiencies, deviations and nonconformances are promptly identified and
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corrected. These measures shall assure that the cause of the condition is
. determined and corrective action is taken to preclude repetition. Section
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17.2.16, " Corrective Action," of the Duke Power Company Quality Assurance
Program Topical Report implements Criterion XVI of 10 CFR 50, Appendix 8.
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The.' licensee's failure to take adequate corrective action for the
violation cited in NRC Inspection Report Nos. 50-369,370/87-05 is
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identified
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10 CFR 50,
Appendix.B,
. Criterion XVI (50-369, 370/87-36-05).
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.6.
Maintenance Observations
Routine ; maintenance activities were reviewed and/o'r witnessed ' by the -
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' resident inspection st'aff'to ascertain procedural and performance adequacy
andlconformance with applicable Technical Specificati ns.
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' The selected activities witnessed were' examinedito ascertain that, where
- applicable,, current written app,oved procedures were available and in use,
that prerequisites were met, that equipment; restoration was completed and
maintenance.results were adequate.
No violation's' or. deviations were identified.
7.
Follow-up of Previous Enforcement Actions-
(Closed)" Violation 50-369/84-23-01 - Failure to fileia special report
regarding . loose parts detection system inoperability.
The licensee's
' Compliance staff, has placed increased emphasis and ef fort' into reviewing :
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' Operations ~ 1ogbooks to' detect -conditions and events that require NRC
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. report's and-notifications. The' resident inspectors have remained vigilant
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for missed reporting requirements and have, from, time to: time, detected
additional failures to submit required reports to the NRC. These failures
have been: attributed to isolated personnel errors and are not. considered
to constitute a programmatic deficiency.
The missed special report was submitted to the NRC on July 30, 1984. On
January 10, 1986, . licensee management issued a Technical Specification
(TS) interpretation to provide guidance to the control room operators in
making operability determinations regarding TS 3.3.3.10, which governs the
-loose parts detection system.
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(Closed) Deviation 50-369/87-14-06 - Failure to construct the auxiliary
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feedwater halon 1301 system in accordance with applicable design drawings
and standards as committed to the'NRC. During a routine tour of the Unit
1-turbine building, the inspector noted that a rack had been installed to-
restrain. the halon cylinders as required
by the appiicable design
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docLments. This item is considered closed.
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(Closed) IFI 50-369,370/87-30-02 - Review the Condensate and ' Feedwater
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operating procedure modification to . prevent auxiliary feedwater system
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automatic starts while performing main feedwater pump trip tests.
The
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applicable procedures (OP/1-2/A/6250/01) and the associated Licensee Event
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Report -(50-370/87-12) were reviewed, and it .was determined that the
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appropriate corrective actions have been implemented. The IFI and the LER
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arr considered closed.
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8,
Preparations for Rufueling
Unit I commencad a refueling outage on September 4.
The inspectors
reviewed /obseryed various ' licensee procedures and preparations to
ascertain the adequacy of the licensee's refueling preparations.
Refueling activities were observed and monitored to ascertain whether
Technical Speci fication requirements were satisfied and activities were
conducted in accordince with approved procedures.
No violations or deviations were identified.
9.
Emergency Preparedness Exercise
The licensee conducted its annual emergency preparedness exercise on
Septembu 11-12, 1987. The NRC resident inspectors and personnel from the
NRC. Region II office participated in the exercise. Details regarding the
exercise will be discussed in NRC Inspection Report Nos. 50-369,370/87-32.
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