ML20056F603
| ML20056F603 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 08/10/1993 |
| From: | Cantrell F, Haag R, Keller L, Wright R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20056F598 | List: |
| References | |
| 50-395-93-19, NUDOCS 9308300144 | |
| Download: ML20056F603 (14) | |
See also: IR 05000395/1993019
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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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ATLANTA, G EORGI A 33323
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YeportNo.: 50-395/93-19
Licensee: South Carolina Electric & Gas Company
Columbia, SC 29218
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Docket No.: 50-395
License No.: NPF-12
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Facility Name:
Virgil C. Summer Nuclear Station
Inspection Conducted:
June 16 through July 31, 1993
Inspectors: MQ C . /M_
g /p /ws
R. C. Haag, Senior Rest' dent Inspector
Date Signed
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L. A. Keller, Resident Inspe% tor
Date Signed
(June 16 through July 9, 1993)
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R. W. Wright, Project Engine'er
Date Signed
(Jul cl2 - 16, 1993)
Approved by:-
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,& Floyd S. Cantrell, Chief
Date Signed
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Reactor Projects Section IB
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Division of Reactor Projects
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SUMMARY
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Scope:
This routine inspection was conducted by the resident inspectors onsite in the
areas of monthly surveillance observations, monthly maintenance observations,
operational safety verification, review of licensee self-assessment
capabilities, onsite follow-up of written reports of nonroutine events at
power reactor facilities, and action on previous inspection findings.
Selected tours were conducted on backshift or weekends.
These tours were
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conducted on six occasions.
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Results:
One violation involving the failure to comply with TS was identified.
This violation dealt with a radiation monitor not being operable when
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containment purge valves were opened. An operator not reviewing the
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applicable procedure contributed to this event (paragraph 5). The efforts to
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resolve an unexpected alarm during a solid state protection system test were
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thorough and well supported (paragraph 3). A decline in the planning and
execution of post maintenance testing was noted for several jobs (paragraph
4). Management involvement and oversight were exhibited in a management
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review board (paragraph 6).
A lack of sufficient information was noted for a
recently issued LER (paragraph 7),
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9308000144 930812
PD9
.\\ DOCK 05000395
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REPORT DETAILS
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1.
Persons Contacted
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Licensee Employees
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W. Baehr, Manager, Health Physics
- C. Bowman, Manager, Maintenance Services
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- M. Browne, Manager, Design Engineering
- B. Christiansen, Manager, Technical Services
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- M. Fowlkes, Manager, Nuclear Licensing & Operating Experience
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- S. Hunt, Manager, Quality Systems
- A. Koon, Nuclear Operations Project Coordinator
- D. Lavigne, General Manager, Nuclear Safety
K. Nettles, General Manager, Station Support
H. O'Quinn, Manager, Nuclear Protection Services
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- M. Quinton, General Manager, Engineering Services
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R. Schwartz, Supervisor, Emergency Planning
J. Skolds, Vice President, Nuclear Operations
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G. Taylor, General Manager, Nuclear Plant Operations
- R. Waselus, Manager, Systems and Performance Engineering
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- R. White, Nuclear Coordinator, South Carolina Public Service Authority
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- B. Williams, Manager, Operations
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Other licensee employees contacted included engineers, technicians,
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operators, mechanics, security force members, and office personnel.
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The following NRC personnel were onsite to observe the annual emergency
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drill and/or tour the plant.
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S. Bajwa, Acting Project Directorate II-1, NRR (July 13-14,1993)
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K. Barr, Emergency Preparedness Section Chief (July 14,1993)
K. Clark, Public Affairs Officer (July 14,1993)
M. Sykes, Intern, NRR (July 12-14,1993)
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R. Trojanowski, State Liaison Officer (July 14,1993)
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G. Wunder, Project Manager, NRR (July 13-14,1993)
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- Attended exit interview
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Acronyms and initialisms used throughout this report are listed in the
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last paragraph.
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2.
Plant Status
The plant operated at or near 100 percent power throughout the
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inspection period.
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Other inspections or meetings:
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On July 14, 1993, the licensee conducted their annual emergency
drill. A regional evaluation team was onsite the week of July 12,
1993, to observe the drill (NRC Inspection Report No. 395/93-20).
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During the week of July 19, 1993, a regional inspection in the areas
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of radiological effluent monitoring and chemistry control was
performed (NRC Inspection Report No. 395/93-27).
3.
Surveillance Observation (61726)
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The inspectors observed surveillance activities of safety related
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systems and components listed below to ascertain that these activities
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were conducted in accordance with license requirements.
The inspectors
verified that required administrative approvals were obtained prior to
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initiating the test, testing was accomplished by qualified personnel in
accordance with an approved test procedure, test instrumentation was
calibrated, and limiting conditions for operation were met. Upon
completion of the test, the inspectors verified that test results
conformed with technical specifications and procedure requirements, any
deficiencies identified during the testing were properly reviewed and
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resolved, and the systems were properly returned to service.
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Specifically, the inspectors witnessed / reviewed portions of the
following test activities:
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a.
Reactor Coolant System (RCS) flow rate measurement (STP 205.002).
This surveillance satisfied the requirement of TS 4.2.3.5 to
determine RCS total flow rate by measurement at least once per 18
months.
The RCS total flow rate was determined by performing a
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mass and energy balance between the primary and secondary system
using basic thermodynamic relationships. While reviewing the test
results the inspector independently verified the accuracy of
several calculations. The total RCS measured flow rate of 289,134
gpm was within the range specified in the Core Operating Limits
Report.
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b.
Turbine drive emergency feedwater pump test (STP 220.002). While
the turbine was running, the inspector noted that the packing on
the governor valve was leaking.
In response to the inspector's
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questions the shift supervisor observing the test stated that while
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some packing leakage was normal (excessive packing force could
interfere with governor valve movement), the observed amount
appeared to be excessive and that a MWR would be initiated. The
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inspector verified that a MWR was generated to correct the leakage.
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c.
Quarterly test of "A" component cooling water (CCW) pump
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(STP 222.002).
Initial condition step 5.1 of the STP requires that
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the CCW system be operating per S0P-118 with either Loop "A" or
Loop "B" as the " active" loop and "C" CCW pump in standby and
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aligned to the active loop. This step appeared to be overly
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restrictive when aligning the system for slow speed and later for
the fast speed testing of the pump. Operations submitted a
procedure feedback form to resolve this issue.
An additional problem was noted while testing the pump in fast
speed. Step 6.3.e of STP 222.002 states to throttle discharge flow
between 10,985 and 11,095 gpm. This corresponded to approximately
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92 percent of full flow.
When flow was being throttled to this
value, the alarms for low flow to the reactor coolant pump thermal
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barriers and bearing coolers were continuously coming in and going
out of alarm condition. This continued for several minutes until
an operator secured CCW flow to the excess letdown heat exchanger.
With additional flow available for the thermal barriers and bearing
coolers, the low flow condition no longer existed. While these
alarms did not adversely impact plant equipment (flow was at the
alarm setpoint and was never lost), they did result in unnecessary
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distraction to the operators. This issue was included on the
procedure feedback form for a procedure change. The inspector
verified that both issues were being pursued by the licensee.
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d.
Engineered safety features actuation system, train "A" slave relay
circuit testing (STP 105.014).
During the portion of the test,
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while CVCS letdown was isolated, the alarm for high letdown relief
valve temperature was received. The alarm setpoint is 160 degrees
Fahrenheit and the operator stated the highest observed temperature
was approximately 200 degrees Fahrenheit. After the test, the
temperature slowly decreased until the alarm cleared. Operators
closely monitored for possible RCS leakage through the relief
valve. The RCS leak rate test that was completed after the slave
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relay test indicated RCS leakage had not increased. Operations
personnel informed the inspector that the relief valve high
temperature annunciator had occasionally alarmed during previous
slave relay tests. The licensee is continuing to review this event
to determine if a change to the test method is warranted.
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e.
Train "A" solid state protection system (SSPS) actuation logic and
master relay test (STP 345.037). On July 23, 1993, during the
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initial portion of the test, the annunciator for intermediate and
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source range trip bypass alarmed while manipulating the multiplexer
test switch.
Since this was an unexpected alarm the test was
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halted.
Engineering and I&C personnel reviewed the system and the
potential cause of the alarm. While they could not determine the
exact cause of the alarm, they were able to verify that the problem
only related to the annunciation function of the SSPS. A
troubleshooting plan was developes 'or use when the SSPS test would
be reperformed the following week. The plan contained repair
instructions, if the alarm recurred, and additional testing for the
repairs.
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The inspector attended the prejob briefing on July 30, 1993, for
the SSPS retest. Management and engineering were actively involved
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in the discussions.
The briefing provided a good overview of the
previous test, subsequent reviews and required actions during the
retest. During the SSPS retest, the same annunciator alarm was
received but at a different sequence in the 5TP
After an
evaluation by I&C and engineering, the action in the
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troubleshooting plan to replace an isolation board was completed.
The remainder of the SSPS test was performed without any additional
problems.
The licensee's efforts to understand the cause and
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resolve the unexpected annunciator alarm were well planned and
'tively supported by engineering and management.
All observed tests were performed in accordance with procedural
requirements and demonstrated acceptable results. Efforts to resolve
an unexpected alarm during SSPS testing were thorough and well
supported by management and other support groups.
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4.
Maintenance Observation (62703)
Station maintenance activities for the safety-related systems and
components listed below were observed to ascertain that they were
conducted in accordance Ath approved procedures, regulatory guides,
and industry codes or sttndards and in conformance.with TS.
The following items w m considered during Gis review:
limiting
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conditions for o p ration were met while components or systems were
removed from service, approvals were obtained prior to initiating the
work, activities were accomplished using approved procedures and were
inspected as applicable, functional testing and/or calibrations were
performed prior to returning components or systems to service,
activities were accomplished by qualified perscr.nel, parts and
materials used were properly certified, and radiological and fire
prevention controls were implemented. Work requests were reviewed to
determine the status of outstanding jobs and to ensure that priority
was assigned to safety-related equipment maintenance that may affect
system performance. The following maintenance activities were
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observed:
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a.
Filter elemeni change for "A" EDG rocker arm lube oil duplex
strainer (PMTS P0166406).
b.
Lubrication activities for the
"C" Component Cooling Water (CCW)
pump and motor (PMTS P0167320, PMTS P0166872). The inspector
verified that the correct amount and type of lube oil was added.
All activities observed were satisfactory.
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c.
dearing replacements for rod drive power supply Motor-Generator
(MG) Set IA (MWR 9304011). On June 29, 1993, operators noted
excessive noise from MG set IA. Subsequent vibration analysis
revealed excessive vibration from the generator inboard bearing.
Inboard and outboard bearings were replaced for both the motor and
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generator. The inspector noted good procedural adherence during
the MG disassembly and bearing replacement activities.
d.
Preventive maintenance for cleaning and inspection of "B"
ground resistor (PMTS P0167657). The inspector observed the
cleaning, inspection of all pertinent ground wiring for damage,
inspcction of terminations for proper tightness, and the
reinstallation of cover bolts. All work was performed
satisfacterily in accordance with EMP 300.005, Cleaning and
Inspection of Electrical Equipment.
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e.
Replacement of "B" EDG temperature thermocouples (MWR 9303494).
The inspector observed the removal and replacement of No. 4 and 6
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cylindar temperature thermocouple indicators which had been reading
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erroneously. The observed thermocouple replacement continuity
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checks were performed satisfactorily. The EDG surveillance testing
conducted subsequently verified the acceptability of these
replacement indicators.
f.
Preventive maintenance (PM) to calibrate the emergency feedwater
pump turbine speed indicator (PMTS P0166113). The PM task sheet
for the calibration specified that no maintenance retest was
required. This was in error since the actual PM task which was
performed per ICP 195.021, Calibrate of Emergency Feedwater (EFW)
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Pump Turbine Speed Indication, also calibrated components in the
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turbine speed control circuit. During a review after completion of
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the work, the licensee recognized that testing of t M turbine speed
control function was required. The error on the PM task sheet for
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retest requirement resulted in an additional run of turbine and
slowed the process for returning the turbine driven EFW pump to
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operable status.
g.
Troubleshooting and repair of the high steam flow alarm and
associated reactor protection system bistable actuation for "B"
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steam generator (MWR 9304151). A summing card in the ste.m
pressure input circuit te tne steam flow comparator had 1 iled and
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resulted in the alr.;; and bistable actuation. A new card was
installed and calibrated, then the circuit was returned to service.
During a review of the work package on the follcwing shift, an I&C
supervisor recognized that e operational test of the steam flow
channel had not been perfo "
follcwing replacement of the summing
card. The channel was the m
out of service and an operational
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test was performed satisfacco- b.
The licensee stated that the
card calibration verified that the instrument channel was operable,
however, their policy is to perform an operational test on a
channel after replacement of component in the circuit. The
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inspector considered that an operational test was a prudent action
following replacement of a circuit card to ensure that the channel
functioned properly.
In General Test Procedure GTP-214, Post
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Maintenance Testing (PMT), the instrumentation test matrix
specifies a channel operation test for circuit card replacements.
Step 4.1.2, of GTP-214, states that a matrix PMT activity, which is
not applicable to th? specific mainter.ance activity, is not
required to be perfs:r cd.
For this maintenance activity, the
inspector considt<<J
w' an operational test was applicable.
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Replacement of a dr a
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control circuit (MWR W A <.i?).
Failure of the card resulted in a
minimum output signal from the chnnel. The operators quickly
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recognized that the channel had 3 iled low and that pressurizer
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heaters had a full demand signa.. Manual pressure control was
selected and plant parameters rema"ned stable. After the card was
replaced and calibrated, the channel was returned to operations for
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c.
functional testing of the control circuit. No instructions were
provided on the method / process for functionally testing the
circuit. The inspector observed the discussions between I&C and
operations concerning restoration of the control circuit, details
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on circuit operation and the best way to test the circuit. The
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inspector considered that these type of issues should be reviewed
prior to performing the PMT and needed information should be
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provided as part of the PMT instructions.
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1.
Troubleshooting the tripping of circuit breaker XMCIDA2Z for "A"
EDG room supply fan (MWR 9303983). A new breaker had been
installed the previcus day when the original breaker failed the
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testing performed by a PM activity. The new breaker then tripped
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when the fan was started. While reviewing this problem, the
electricians noted that the replacement breaker had been set to
trip at the low end of the trip tolerance band. After the breaker
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was reset at the higher end of the band, the breaker and fan
operated satisfactorily.
In a subsequent conversation, the
inspector was informed by engineering that the tripping of a
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breaker, when set at the low end of the trip tolerance, is not
unexpected. The normal corrective action would be to reset the
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breaker at the higher end of the band and determine if the
equipment will operate.
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Tightening the hinge pin cover bolt for turbine driven EFW pump
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discharge check valve XVC1016 (MWR 9303976).
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Retorquing the body to bonnet fasteners for "A" charging pump
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suction valve XVG8471A (MWR 93N3136).
Boric acid crystals on the
valve indicated that previous leakage had occurred, however, no
current leakage was noted.
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Replacement cf voltay regulator raise / lower switch located on the
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MCB for "B" EDG (MWR 9303394). This work was started on June 21,
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1993.
Previously on June 3,1993, during an EDG surveillance test,
the output breaker tripped open due to a actuation of the
overcurrent relay. The overcurrent condition was attributed to
sticking of the voltage regulator switch,
i.e., not returning on
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its own to the neutral position.
In response to the problem with
the sticking switch, operations issued a station instruction (SI)
to inform operators of the problems experienced with the switch and
actions to take if the EDG voltage needed to be adjusted. The
inspector reviewed the SI and concluded that the appropriate
information had been provided to the operator. However, the
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inspector noted that the addition of a caution tag on the switch
would have provided a more positive means of informing operators of
the deficiency. This type of visual aid would be important to an
operator during an event response if manual control of EDG voltage
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was required.
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The maintenance activities observed were well executed.
Problems
were noted with incomplete specification of post maintenance
testing requirements and the lack of detail for the specified
testing.
5.
Operational Safety Verification (71707)
a.
Plant Tour and Observations
The inspectors conducted daily inspections in the following areas:
control room staffing, access, and operator behavior; operator
adherence to approved procedures, TS, and limi. ting conditions for
operations; and review of control room operator logs, operating
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orders, plant deviation reports, tagout logs, and tags on
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components to verify compliance with approved procedures.
The inspectors conducted weekly operability inspections of
selected ESF systems for valve alignment, breaker positions,
condition of equipment or component (s), and operability of
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instrumentation and support items essential to system actuation or
performance. The emergency borate mode of the reactor makeup
system, the containment spray system, and the diesel fuel oil
trar,sfer system were included in these inspections.
P' ant tours included observation of general plant / equipment
conditions, fire protection and preventative measures,
radiation
pratection controls, physical security controls, and plant
hcusekeeping conditions / cleanliness.
Reacter coolant system leak
rates were reviewed to ensure that detected or suspected leakage
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from the system was recorded, investigated, and evaluated; and
that appropriate actions were taken if required.
Selected toun
were conducted on backshifts or weekends.
b.
Reactor Building (RB) Alternate Purge Supply Valves Opened While
RB Radiation Monitor Was Out of Service
Technical Specification (TS) 3.6.1.4 requires that RB internal
pressure be maintained between -0.1 and 1.5 psig. Occasionally RB
pressure drops when the ambient outside air is relatively cool.
When this occurs, it is necessary to increase RB pressure via the
RB purge intake fan (XFN-95) which draws in outside air and
discharges into containment through the purge valves.
System
Operating Procedure SOP-ll4, Reactor Building Ventilation System,
Section III N, Reactor Building Normal Pressure Control, is the
appropriate procedure for this activity.
Initial condition,
step 1.8 of Section III N, requires that RB atmosphere gaseous
monitor, RM-A2, be in service for the normal RB pressure control
operations.
This 9rocedural requirement is intended to ensure
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compliance with TS Table 3.3-3, Item 3.c.2, which requires the
containment purge supply and exhaust valves be maintained closed
if RM-A2 is out of service.
On June 28, 1993, at 0400, RM-A2 was tagged out for routine
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preventative maintenance. The fact that RM-A2 was out of service
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for PM was discussed during the following shift turnover meeting
held at 0715. At approximately 0745, in response to a RB low
pressure alarm, the control building operator opened the alternate
purge supply valves (PVG-6056, 6057) and started XFN-95 in order
to increase RB pressure, despite RM-A2 being out of service. The
control room was not informed prior te the operator starting these
activities.
During subsequent reviews, the inspector discovered
that the control room would not typically be informed prior to
starting this activity or similar tasks.
In a subsequent
interview, the operator stated that 50P-114 was not looked
at/ reviewed prior to this evolution. The fact that the RB
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pressure is routinely controlled by performing these actions was
the primary reason stated by the operator for the S0P not being
reviewed. Several minutes after the opening of the purge valves,
a control room operator recognized this condition and the valves
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were subsequently closed. The failure to ensure RM-A2 was
operable prior to opening a containment purge valve is a violation
of TS and is identified as Violation 395/93-19-01.
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In 1990, the licensee initiated efforts to provide additional
guidance on the usage and adherence of procedural instructions.
Station Administrative Procedure SAP-123, Procedure Use and
Adherence, was issued on June 20, 1991, to provide this type of
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guidance.
The primary method for specifying how to use a
procedure is to stamp the " usage level" on the procedure cover
sheet.
The cover sheet for 50P-Il4 contains the following
procedure usage guidance; ' Multiple Usage Levels: This Procedure
Contains Multiple Usage Levels.
Refer to the. Procedure Content
for Clarification". However, no additional guidance is provided
in the 50P on usage levels. SAP-123 requires that each procedures
stamped " Multiple Usage" have the appropriate usage level clearly
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described within the procedure.
Yet, in the step 6.6 " Program
Implementation", of SAP-123, the date for incorporating procedural
usage requirements into " Multiple Use" procedures is September 30,
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1993. The inspector considered that the greater than two year
time period for incorporating the usage levels into plant
operating procedures was excessive.
Since 50P-114 did not contain the specific usage levels,
managements expectations on the usage of SOP-Il4 had not been
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conveyed to the procedures users. Earlier, on June 7,1993, an
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operator error rendered safeguards equipment inoperable.
In this
event the applicable procedure had not been reviewed for the
activity involving the personnel error (NRC Inspection Report 50-
395/93-17).
In reviewing the licensee's response to the earlier
event, the inspector could not find any action related to
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improvements in procedure usage. The inspector is concerned that
the licensee's corrective actions have not prevented these type
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errors.
c.
Change to Emergency Feedwater TS Surveillance Requirement
While reviewing the conditions required for testing the turbine
driven emergency feedwater pump (TDEFP), the licensee recognized
that the initial test condition could:not be met. TS 4.7.1.2.a.2
required that the TDEFP be tested at taast once per 31 days when
secondary steam supply pressure is greater than 900 psig. The
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TDEFP was tested satisfactorily on April 29, 1993, after the
completion of the seventh refueling outage while the plant was in
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mode three. However, due to steam generator tube plugging during
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the outage the secondary steam pressure while at power was less
than 900 psig.
On May 25, 1993, the licensee request discretion not to enforce
compliance with the required action of TS 4.7.1.2.a.2 based on the
TDEFP being able to develop the required flow and differential
pressure at a lower steam pressure. On May 27, 1993, NRR issued a
hotice of Enforcement Discretion for testing the TDEFP at steam
pressure greater than or equal to 865 psig. The licensee
submitted a proposed TS change to incorporate'the 865 psig steam
pressure requirement in lieu of the 900 psig requirement.
6.
Review of Licensee Self-Assessment Capability (40500)
On July 20, 1993, the inspector attendsd a management review board
(MRB) which discussed sevcral recent issues involving failure to follow
procedures and personnel errors. Several of the issues had been
discussed in an MRB held the previous week, and in this HRB additional
information was presented. The d!scussions of these events were open
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and focused on the need to better understand the events and
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contributing factors to the events. Some of the additional action
items and corrective action addressed by the MRB related to general
problem areas involving personnel errors and inattention to detail
which were common to many of the events.
7.
Onsite follow-up of Written Reports of Ncnroutine Events at Power
Reactor Facilities (92700)
(Closed)
LER 92-02, Improper Operation of Containment Isolation Valve
Due to Personnel Error
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On January 28, 1992, an operations review of inoperable equipment logs
found that on three separate occasions containment isolation valve XVA-
9312B, " Sample Line Return From Radiation Monitor", had been
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temporarily opened to facilitate the performance of a required
surveillance activity. Operations had previously failed to recognize
that these temporary restorations of the inoperable valve were in
conflict with the intent of TS 3.6.4, Containment Isolation Valves and
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plant procedures. Temocrary restorations of this nature are allowed
only for troubleshooting, maintenance, and surveillance testing of the
component to prove operability.
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The faulty limit switch positioner on isolation valve, XVA-9312B, was
repaired and the valve was returned to operable status on February 8,
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1992. Revision 7, Change A, to SAP 205, Status Control and Removal was
reviewed by the inspector and found to further enhance and clarify
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instructions in the area of temporary restoration to service for
equipment. The inspector examined licensee response form CGSS 01-4652-
N0 dated March 31, 1992, which documents that the , subject event and
specifically the failure to comply with the requirements of SAP 205 had
been reviewed by all operations personnel. The above actions are
considered satisfactory and should be adequate to prevent a recurrence
of this event.
(Closed)
LER 92-04, Reactor Trip Due to Low-Low Level In "C" Steam
Generator
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The reactor protection system responded as designed.
A post trip
review revealed that six condenser steam dump valves (Steam Dump Bank
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2) did not fully open on demand.
This was caused by the earlier
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replacement of " current to pneumatic" transducers (Rosemount) that had
higher impedance than the original (Fisher) transducers. This higher
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impedance prevented the power supply card for Steam Dump Bank 2 from
developing a full open signal of 20 milliampere. Consequently the
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valves would not fully open on a 100 percent demand signal.
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A search of other installed Rosemount transducers confirmed that no
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more than three valves are being driven by a single power supply card
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which was deemed adequate. The power supply cards for steam dump valve
transducers were rated to apply full load current at an impedance of
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1250 ohms.
The impedance of the Fisher transducers is 176 ohms versus
410 ohms maximum for Rosemount transducers.
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The transducers for the steam dump valves were replaced with the
original lower impedance transducers and their calibration frequency
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was increased to counteract their tendency to drift.
Since these
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Rosemount transducer replacements were evaluated and approved under the
licensee's, Equal To/Better Than (ET/BT) program, the subject procedure
was reviewed to determine if there were any needed enhancements.
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Review of the ET/8T procedure ES-419, Revision 5, revealed it was
!
revised to caution the engineer to consider the potential effect of
multiple applications / installations to assure performance requirements
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can be maintained. Additionally, engineers were cautioned to
understand what post installation testing will occur and to determine
if additional testing / acceptance criteria should be specifically
required to assure performance or functional characteristics.
(Closed)
LER 92-06, Substantial Safety Hazard (Part 21) Report on
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f eedwater Isolation Valve (FWIV) Actuators
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The subject LER dated July 31, 1992, was submitted to the NRC pursuant
to 10 CFR 21.21(c)(3)(ii) and 10 CFR 50.73 requirements.
During the sixth refueling outage, a plant modification was implemented
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to replace the original Anchor Darling hydraulic-pneumatic FWIV
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actuators with Chicago Fluid Power (CFP) actuators. On November 13,
1991,
"C" FWIV failed during post modification testing due to a faulty
poppet seal in the control valve assembly. Additional failures
occurred during extended testing of the actuator. This resulted in a
1
decision to replace all the poppet seals with seals of the same design.
Due to these failures, the manufacturer suggested that each new poppet
seal be tested prior to installation by twisting and bending the seal
by hand. Several new unused seals failed this inspection due to cracks
developing in the corner similar to the earlier failures.
Finite
element analysis performed on the poppet seal design disclosed that the
maximum calculated tensile stress to be 2668 psi versus an allowable of
1400 psi.
In conclusion, the old poppet seals were determined to be
inadequately designed for use on the new CFP FWIV actuators. A new
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seal design was developed using the same material but utilizing _
thicker lip and an increased inside corner transition radius. ine
poppet seals in all FWIV control assemblies were replaced with new
designed seals and the valves were tested satisfactorily.
(0 pen)
LER 93-04, Missed TS Surveillance Due to Personnel Error
This event involved a failure to document a test deficiency and the
failure of the review process to recognize that the test had not been
performed since the required signoff had not been completed. The
component was a secondary isolation valve for reactor building spray
sump "B" and the test involved stroke testing via the associated ESF
relay. However, the LER lacked sufficient detail to fully understand
the event and it's significance to determine if the specified
corrective action was adequate.
The inspector informed the licensee
that LER 93-04 lacked sufficient detail to allow review and final
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closure of the LER. This LER will remain open pending additional
information being provided on this event.
8.
Action on Previous Inspection Findings (92701, 92702)
(Closed) Violation 395/91-17-01, Personnel Error Resulting in an ESF
Actuation and LER 91-06.
An inadvertent start of "B" EDG occurred when safeguards system
actuation testing was not performed in the sequence specified by the
surveil'ance test procedure (STP).
In the restoration section of the
STP which generates an emergency start signal for the EDG, there was a
step to reset the start signal. However, the EDG was returned to
service before the signal was reset. The involved operator was
counselled on the importance of attention to detail and procedural
compliance.
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When this 18 month test was previously performed, a procedural error
caused an inadvertent start of the "B" EDG. After review of these and
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similar events, the licensee initiated action to strengthen the program
for infrequently performed tests and activities.
Procedural changes
were made to define these type of activities and to specify the program
improvements. The inspectors have noticed increased attentior, being
provided to infrequently performed activities.
Prejob briefing and
management involvement have been the key areas of this increased
attention.
!
(Closed) Violation 395/91-23-03, Inadequate Hydrostatic Test for a
Piping Modification
This violation dealt with a hydrostatic test performed on newly
installed chill water system piping.
The test failed to pressurize all
applicable piping sections. After the error was recognized, the
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hydrostatic test that was reperformed included those portions of the
modified system that had not been previously tested. An engineering
review of actual post modification instructions was added via a
procedure change to SAP 600, Application, Scheduling, and Handling of
Maintenance Activities. This is to ensure that the intended scope of
testing is actually transformed into the maintenance / test instructions.
In addition, engineering and test personnel received training on
lessons learned from this event and reemphasis on their
responsibilities involving post modification testing.
9.
Exit Interview (30703)
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The inspection scope and findings were summarized on August 2, 1993,
with those persons indicated in paragraph 1.
The inspectors described
the areas inspected and discussed the inspection findings.
No dissenting comments were received from the licensee. The licensee
did not identify as proprietary any of the materials provided to or
,
reviewed by the inspectors during the inspection.
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Item Number
Description and Reference
395/93-19-01
Violation - Failure to follow procedural
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instructions resulted in the noncompliance
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of TS.
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10.
Acronyms and Initialisms
Component Cooling Water
CFP
Chicago Fluid Power
Chargirg and Volume Control System
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Emergency Feedwater
EMP
Electrical Maintenance Procedure
Engineering Services Procedure
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Engineered Safety Feature
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ET/BT
Equal To/Better Than
1
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Feedwater Isolation Valve
GPM
Gallons Per Minute
GTP
General Test Procedure
Instrumentation and Control
Instrumentation Control Procedure
LER
Licensee Event Report
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Hotor-Generator
Hanagement Review Board
MWR
Maintenance Work Request
NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
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Preventive Maintenance
PMTS
Preventive Maintenance Task Sheet
Pounds Per Square Inch
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Pounds Per Squaro Inch Gauge
Reactor Building
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Radiation Work Permit
Station Administrative Procedure
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Station Instruction
S0P
System Operating Procedure
SPR
Special Report
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SSPS
Solid State Protection System
Surveillance Test Procedure
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TDEFP
Turbine Driven Emergency Feedwater Pump
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TS
Technical Specification
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A