ML20246B026
| ML20246B026 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/25/1989 |
| From: | Cantrell F, Maxwell G, Miller W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20246B007 | List: |
| References | |
| 50-348-89-11, 50-364-89-11, IEIN-87-050, IEIN-87-50, IEIN-89-033, IEIN-89-044, IEIN-89-33, IEIN-89-44, NUDOCS 8907070175 | |
| Download: ML20246B026 (15) | |
See also: IR 05000348/1989011
Text
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UNITED STATES
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NUCLEAR REGULATORY' COMMISSION
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.101 MARIETTA ST., N.W.
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ATLANTA, GEORGIA 30323
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LReport Nos.: 50-348/89-11 and 50-364/89-11
Licensee:
Alab'ama Power Company
,
600 North 18th Street:
Birmingham, AL 36291
Docket Nos.:
50-348 and 50-364
Facility Name:
Farley 1 and 2
Inspection Conducted:' April 11'through May 10.-1989
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Inspec rs: .
ell, L
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G.~F. Ma
Seni o Resident Inspector-
Date Signed
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W. 'H. ,M;11et, Jr. , Resident Inspector
Date Signed
Approved by: ~
[!2J/b
F.'S.Cantrell,6p)fonChief
Cate Sfgned
Division of Reactor Projects
SUMMARY
Scope:
This routine onsite inspection . involved a review of operational safety
verification, monthly surveillance
observation, monthly maintenance
observation, licensee. event reports, Unit 2 startup from refueling, response to
NRC information notices, followup of written reports (Part 21) and previous
inspection findings.
Results:
N
Within .the areas inspected, the following violations were identified: Failure
to maintain containment integrity, paragraph 3.b.(1), and Failure to follow
procedure while preparing the solid state protective system for a surveillance
-test, paragraph 3.b.(2).
The licensee was requested to review his practice of
leaving high-low interface valves energized.. paragraph 8.c.
Certain tours were conducted on deep backshift or weekends, these tours were
conducted on April 21, 30 and May 7 (deep backshift inspections occur between
10 p.m. and 5 a.m.).
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8907070175 890525
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ADOCK 05000348
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REPORT. DETAILS
- 1.
Licensee Employees Contacted
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R. G. Berryhill, Systems Performance and Planning Manager
C. L. Buck, Plant Modification Manager
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L. W. Enfinger, Administrative Managet
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R. D. Hill, Assistant General Manager - Plant Operations
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' D. N.- Morey, General Manager - Farley. Nuclear Plant
C. D. Nesbitt, Technical Manager-
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J. K. Osterholtz, Operations Manager
L. M. Stinson, Assistant General Manager - Plant Support
J. J.' Thomas, Maintenance Manager
L. S. Williams, Training Manager
Other licensee. employees contacted included, technicians, operations
personnel, maintenance and I&C personnel, security -force members, and-
- ffice personnel.
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The NRC Project Section Chief, F. S. Cantrell, visited the Farley. site
May 9-11, 1989, to tour .the site and meet with licensee management and
the resident inspectors.
' Acronyms and abbreviations used throughout this report are . listed in the
'last paragraph.
2.
Plant Status
Unit 1
Unit 1 operated at approximately 100% reactor power throughout the
reporting period.
Unit 2
Unit 2 was shutdown throughout the reporting period for a normal scheduled
refueling outage.
3.
Operational Safety Verification (71707)
a.
Plant Tours
The inspectors conducted routine plant tours during this inspection
period to verify that the license requirements and commitments were
being implemented. Inspections were conducted at various times
including week-days, nights, weekends and holidays. These tours were
performed to verify that: systems, valves, and breakers required for
safe plant operations were in their correct position; fire protection
equipment, spare equipment and materials were being maintained and
stored properly; plant operators were aware of the current plant
status; plant operations personnel were documenting the status of
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out-of-service equipment; there were no undocumented cases of unusual
fluid leaks, piping vibration, abnormal hanger or seismic restraint
movements; all reviewed equipment requiring calibration was current;
and, general housekeeping was satisfactory.
Tours of the plant included review of site documentation and
interviews with plant personnel. The inspectors reviewed the control
room operators' logs, tag out logs, chemistry and health physics
logs, and control boards and panels.
During these tours the
inspectors noted that the operators appeared to be alert, aware of
changing plant conditions and manipulated plant controls properly.
The inspectors evaluated operations shift turnovers and attended.
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shift briefings.
They observed that the briefings and turnovers
provided sufficient detail for the next shift crew and verified that
the staffing met the TS requirements.
Site security was evaluated by observing personnel in the protected
and vital areas to ensure that these persons had the proper
authorization to be in the respective areas.
The inspectors also
verified that vital area portals were kept locked and alarmed.
The
security personnel appeared to be alert and attentive to their duties
and those officers performing personnel and vehicular searches were
thorough and systematic.
Responses to security alarm conditions
appeared to be prompt and adequate.
Selected activities of the licensee's Radiological Protection Program
were reviewed by the inspectors to verify 'conformance with plant
procedures and NRC regulatory requirement.
The areas reviewed
included: operation and management of the plant's health physics
staff, "ALARA" implementation, Radiation Work Permits (RWPs) for
compliance to plant procedures, personnel exposure records,
observation of work and personnel in radiation areas to verify
compliance to radiation protection procedures, and control of
radioactive materials,
b.
Plant Events and Observations
(1) Loss of Unit 2 Containment Integrity
On April 19, at approximately 12:15 p.m., Unit 2 containment
integrity was breached during fuel movement when the bonnet to
valve Q2P25V001A in the 1/2 inch chemical injection line to
steam generator line 2A was removed with the manways and
handhole covers for steam generator 2A also removed.
This
created an air to air path from the containment to the main
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steam valve room, which is located outside of the containment
structure, through the open valve in the chemical injection
piping system.
The licensee estimated that the valve bonnet was
removed and containment integrity was breached for approximately
30 minutes.
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' Work order MWR 182030, was issued and released by the operations
group to repack valve .Q2P25V001A.
The maintenance workers
assigned to the repacking job found the valve-bonnet bolts to be
only finger tight.
To eliminate the valve leak the workers
removed the valve bonnet, replaced the gasket and reinstalled
the bonnet.
However, no approval was obtained to deviate from
the original work scope.
On April 20, during a review of
completed work raquests, the licensee identified that this
maintenance work had resulted in a breach of containment
integrity.
Initial investigation indicated that this breach had
existed for appro/ir.ately 30 minutes (between 12:15 and
12:45 p.m.).
Furt kr investigation by the Licensee
(LER364/89-04) ind!cated that the packing for the valve
was removed at ap,)roximately 9:00 a.m. on April 19.
All work
was completed on the valve by 4:00 p.m. that day.
Containment
integrity was breached several times during this period;
however, the opening which resulted from removal of the packing
only is considered inconsequential with the bonnet removed, the
opening is more significant.
TS Section 3.9.4 states that during refueling operations each
containment building penetration providing direct access from
the containment atmosphere to the outside atmosphere will be
closed by an isolation valve, blind flange or manual va? a.
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This open penetration is identified as Violation 364/89-11-01,
Breach of Containment Integrity During Refueling.
The severity of this violation is reduced due to: the small size
of the open penetration (1/2 inch); short time of open
penetration; and, the temperature and pressure within the
containment being approximately the same as that which existed
outside of containment at the time of the event.
Even considering
a fuel handling accident, the safety significance is minimal.
(2) Unit 2 Inadvertent Safety Injection
On April 29, while licensee personnel were conducting the
preliminary steps for a surveillance test the plant experienced
a false start signal for the
"A"
train safety injection
equipment. When the event occurred the plant was in Mode 5 (cold
shutdown). The " A" train safety injection equipment which
started, included emergency diesel generator 1/2 A, HSSI, RHR,
The inspectors reviewed plant logs and documentation and
interviewed operators, supervisors and plant management
personnel concerning the event.
It was determined that the
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following sequence of events occurred prior to, during or after
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the false start:
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Operations personnel were in the process of establishing-
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the initial. conditions to allow surveillance test procedure
FNP-2-STP 16.6, Spray and Phase B Actuation Test, Revision 9
to be conducted.
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The. STP would eventually require re-positioning the solid
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state protection system SSPS mode selector switches.
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To allow the manipulation of the mnde selector switches the
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operators were required to first obtain a " partial release"
for the removal of clearance tags which were previously
.placed on the switches by maintenance instrumentation and
controls personnel.
The clearance tags had been , installed as a part of.-
procedure. IMP-0.7, Modes 5 and 6 Surveillance Test
Performance. This procedure was implemented earlier during
the outage-to disable solid state protection system output
relays during the outage, to prevent spurious activations.
FNP-2-STP-16.6, Section 3.9 states "If IMP-0.7 is in effect
in the SSPS, then perform the.following before running this
STP...in .the A Train SSPS logic cabinet place the input
error inhibit switch to the inhibit position... then ... in
' the A Train SSPS output cabinet place the mode selector
switch to the operate position."
Contrary to the above requirements of 2-STP-16.6 Section 3.9,
the operators, placed the SSPS mode selector switches
in the operate position while removing the clearance tags,.
without first placing the input error inhibit switches in
the inhibit position. As a result a SI signal was' generated.
The operators who were stationed at the main control boards
immediately recognized that a false start signal had been
generated.
They then took all of the required procedural
steps to return the plant to its' normal mode 5
configuration.
The shift supervisor then reported the event to the NRC
duty officer.
The above discrepancy is identified as Violation 364/89-11-02,
Failure to follow procedure while preparing the SSPS for a
surveillance test.
(3)
Filling and Venting Unit 2 RCS
The inspectors observed portions of the filling and venting
operations after the Unit 2 reactor vessel head was reinstalled.
The plant operators followed procedure 2. 50P-1.2, RCS Filling
and Venting, to satisfactorily accomplish these evolutions.
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(4).RadiationIncidentReportNo.89-006'
During a routine plant tour on May 7, the inspectors noted a
system operator enter RHR pump room 2B which was posted as a
"High Radiation" area. The operator entered this area without
either an alarming digital dosimeter or a HP technician with a
dose rate survey instrument.
Radiation Work Permit 0-89-0003
posted at the door to this pump room required an alarming
digital dosimeter or a HP technician with a dose rate survey
instrument.
The system operator apparently violated the RWP.
For additional' information on this item refer to NRC Report
348,364/89-13.
Except as noted, no violations or deviations were identified. The results
of the inspections in this area indicate that the program was effective
with respect to meeting the safety objectives.
4.
Monthly Surveillance Observation (61726)
The inspectors witnessed the licensee conducting maintenance surveillance
test activities on safety-related systems and components to verify that
the licensee performed the activities in accordance with TS and licensee
requirements. These observations included witnessing selected portions of
each surveillance, review of the surveillance procedures to ensure that
administrative controls and tagging procedures were in force, determining
that approval was obtained prior to conducting the surveillance test and
that individuals conducting the test were qualified in accordance with
plant-approved procedures.
Other observations included ascertaining that
test instrumentation 'used was calibrated, data collected was within the
specified requirements of TS, any identified discrepancies were properly
noted, and the systems were correctly returned to service. The following
specific activities were observed:
1-STP-9.0
Reactor Cooiant System Leak Rate Test
1-STP-15.0
Containment Air Lock Door Seal Operability Test
2-STP-18.3
Containment Purge and Exhaust Valve IST
2-STP-18.5
Containment Mini-Purge and Exhaust Valve IST
2-STP-34
Containment Inspection
2-STP-40.0
Safety Injection With Loss of Off-Site Power
2-STP-40.5
Charging /HHSI Pump 2A, 2B & 2C Low Discharge Head Flow Test
0-STP-80.1
Diesel Generator 1-2A Operability Test
0-STP-80.2
Diesel Generator IC Operability Test
2-STP-108
Incore Thermocouple
2-STP-131.07
Smoke Detector Function Test (Containment)
2-STP-160.23
Train "B" RHR Suction Line Hydrostatic Test
1-STP-754
Verify R-23A/23B in service once per 24 Hours when R-60 is
2-STP-905
Auxiliary Building Battery Inspection
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0-ETP-3616
Performing Monthly Flux Maps (Data Collection for STP-108,
110 and 121)
0-ETP-3640
Vessel TV Inspection Below Lower Core Plate (Unit 2)
0-ETP-3643
Verification of Rod Control System Operability (Unit 2)
a.
Procedure 2-STP-40.0
Surveillance 2-STP-40.0 was conducted on April 21, but was not
satisfactory.
Several valves and components did not function
properly and inverter 20 failed.
This prevented resetting "B" train
following initiation of the safety injection signal. The items which
failed were retested under other procedures to verify operability.
b.
Procedure 2-STP-905
During the evaluation of the electrical maintenance activities being
conducted on battery 2B for surveillance 2-STP-905, the inspectors
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noted that the isolation breaker for the battery was not tagged open.
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The controlling work authorization (WR-83535) and 2-STP-905 were
reviewed and found not to contain any specific requirements to tag
the isolation breaker (LB18-Q2R42E002B-B) open. The STP did contain
requirements that the breaker be placed in the open position and that
the battery cables, once disconnected, be taped with electrical
insulation tape on their exposed ends. The inspectors discussed the
above as found condition with the electrical maintenance personnel
who were conducting the work and also with the electrical maintenance
supervisor.
As a result of the inspectors concern about the safety
of those personnel who conduct this STP, the electrical isolation
breaker LB18 was promptly tagged open.
The licensee is currently
evaluating the site practices concerning tagging requirements for
work involving low voltage power sources such as the plant batteries.
c.
Procedure 1-STP-'754
The radiation monitor for the main steam relief and atmospheric steam
dump discharge from loop 1A monitor RE-60A, was declared inoperable
on November 30, 1988.
The licensee reported this to the NRC by LER
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88-23 and initiated the preplanned alternate method of monitoring
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the main steam relief and atmospheric steam dump discharge.
This
method included verification that radiation monitors RE-23A and 23B
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on the steam generator blowdown system were in service once per 24
hours or conducting an appropriate grab sample if monitors RE-23A
and 23B were not in service.
These alternate methods are
accomplished and documented by procedure FNP-1-STP-754, RE-60A, B, &
C Contingency Sampling.
On April 14 and 17, the inspectors reviewed
recently' completed surveillance
of 1-STP-754 to verify that
appropriate measures were being implemented while radiation monitor
RE-60A was out of service.
During this review it was noted that the
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steam generator blowdown ' system for Unit I was out of service from
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April 5 at 11:03 a.m. until April 7 at 5:15 p.m.
On these dates
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radiation monitors RE-23A and 23B were not operable since there was
no flow through the steam generator blowdown system.
were not taken from 'the steam generator outlets as required when
RE-23A and 23B are not in service.
This problem was discussed with the plant chemistry group and it was
determined that the cause of the problem was a procedure inadequacy.
The licensee noted that the use of radiation monitor RE-15A on the
steam jet air ejector would provide an improved method of monitoring
parameters when RE-60A, B or C are out of service.
would be required if RE-15A was not in service.
Procedure
FNP-1-STP-754 was promptly revised to incorporate this change.
The
failure to implement the previously approved preplanned compensatory
measures while RE-60A and RE-23A and 23B were out of service on April
6 and 7 is a procedure violation. This violation has minimal safety
significance since monitor RE-15A was available to detect any
radiation within the secondary system, and is not cited because the
criteria specified in Section V. A of- the enforcement policy were
satisfied, NCV 348/89-11-02.
Excepted as noted, no violations or deviations were identified.
the
results of the inspections in this area indicate that the prooram was
effective with respect to meeting the safety objectives.
5.
Monthly Maintenance Observation (62703)
The inspectors reviewed the licensee's mainten.ince activities to verify
the following: maintenance personnel were obt.ining the appropriate tag
out and clearance approvals prior to commencing work activities, correct
documentation was available for all requested parts and material prior to
use, procedures were available for all requested parts and material prior
to use, procedures were available and adequate for the work being
conducted, maintenance personnel performing work activities were qualified
to accomplish these tasks, no maintenance activities reviewed were
violatirig any limiting conditions for operation during the specific
evolution, the required QA/QC reviews and QC hold points were implemented,
post-maintenance testing activities were completed, and equipment was
properly returned to service after the completion of work activities.
Activities reviewed included:
MWR-164766
Cut and cap Unit 2 BIT bypass line.
MWR-182604
Overload test of Unit 1 MOV347BA, outside air to control
room filter motor starter.
MWR-183040
Disassemble / reassemble Unit 2 MSIV Q2N11V001B for
preventive maintenance using Procedure 0-MP-39-0.
MWR-183120
Replace spring for safety relief valve on service water to
RHR 2B room cooler and set relief at 154 psi.
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MWR-193022
Inspection of fuses in Unit 1 panel cabinet J
(Q1HINGB325047) for B0P control panel.
MWR-195302
Replace "PCB" in transformer 2F with new dielectric
solution (Procedure EMP-3573.02).
MWR-197064
Inspect and rebuild Unit 2 snubber 2 SI-R997A for safety
injection line.
MWR-198441
Repairs to Unit 2 electrical penetration Q2T52B040.
WA-WOO 302462
Statior service transformer 2F supply breaker inspection,
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test and maintenance using Procedure EMP-1313.03.
2-PMP-1078
Hydrostatic test of Train A service water header (0TC
890427-1).
0-MP-89.0
Limitorque motor operated valve testing using "M0 VATS"
testing system (valve No. Q2N21M0V3232C).
On April 28, the inspectors witnessed a hydrostatic test of Unit 2 service
water train B header which was unsatisfactory. The hydrostatic test pumps
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were unable to maintain the required test pressure for 10 minutes.
Subsequently, the test procedure was revised to permit the system to be
pressurized by the train B system through cross train interconnecting
piping. This portion of the test was not observed by the inspectors.
No violations or deviations were identified.
The results of the
inspections in this area indicate that the program was effective with
respect to meetina the safety objectives.
6.
Licensee Event Reports (92700, 90714)
The following Licensee Event Reports (LER) were reviewed for potential
generic problems to determine trends, to determine whether information
included in the report meets the NRC reporting requirements and to
consider whether the corrective action discussed in the report appears
appropriate.
Licensee actions were reviewed to verify that the event
has been reviewed and evaluated by the licensee as required by the
Technical Specifications; that corrective action was taken by the
licensee; and that safety limits, limiting safety settings and LCOs were
not exceeded.
The inspector examined the incident reports, logs and
records, and interviewed selectcu a vrsonnel.
The following reports are
considered closed:
Unit 1
LER/88-18
Failure to adequately eddress effects of fire in an area.
Unit 2
LER/88-02
TS 3.0.3 entered when a fire damper in the penetration room
filtration system common suction line was closed.
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LER/88-06
Special Report: Fire detection system inoperable for more
than fourteen days.
LER/89-03
Steam generator tube plugging.
No violations or deviations were identified.
The results of the
inspections in this area ' indicate that the program was effective with
respect to meeting the safety objectives.
7.
Unit 2 Plant Startup from Refueling (71711)
The licensee implemented administrative procedures to assure that systems
disturbed or tested during the refueling outage were returned to operable
status before plant startup.
Procedure FNP-0-AP-16, Conduct of
Operation - Operations Group, Appendix C, Return to Service and System
Lineup, requires a return to service check list to be prepared to insure
that all necessary work and actions tre completed prior to returning the
plant to service.
Procedures FNP-2~dOP-1.1B, Startup of Unit from Cold
Shutdown to Hot Standby, and FNP-2-dOP-1.2, Startup from Hot Standby to
Minimum Load, require additional verification of the completion of system
lineups and completion of surveillance test procedures before entrance
into the next appropriate operational mode.
The inspectors reviewed the
check lists required by these procedures and verified that the licensee's
program was effective.
Completed system lineup check list procedures were
reviewed, to verify that the restoration program was being implemented.
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No discrepancies were noted.
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On the evening shift of May 7, the inspectors performed a post refueling
outage inspection of the Unit 2 containment to verify that the general
cleanliness was acceptable for plant entry into Mode 4.
Practically all
of the tools, equipment and materials which had been moved into containment
to support the refueling outage had been removed. The inspectors reviewed
the deficiency check list o/ procedure 2-STP-34, Containment Inspection,
and noted that NRC identified discrepancies were also on the licensee's
checklist. These items included: tool boxes, drain hoses, HP supplies and
several miscellaneous areas which needed additional cleaning.
The
licensee stated that all of these discrepancies were corrected prior to
entry into Mode 4.
Accessible key portions of the Unit 2 containment spray system and
component cooling water system were inspected by the inspectors to verify
that: valves and electrical breakers were in correct alignment; hangers
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and supports were properly made up; and major components were labeled,
lubricated, cooled and no visible leakage existed.
Portions of these
systems had been disturbed during this outage, but based on this
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inspection, these systems appears to have been returned to service in
accordance with the applicable procedures.
However, several minor
discrepancies were identifieo, reported to the licensee, and were promptly
corrected.
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The inspectors witnessed portions of the reactor thermocouple RTD cross
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calibration tests (2-STP-198) and rod drive control system operability
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tests (0-ETP-3643) which were conducted before the reactor coolant systems
were considered ready for operation.
No violations or deviations were identified.
The results of the
inspections in this area indicate that the program was effective with
respect to meeting the safety objectives.
8.
Review of Licensee's Response to NRC Information Notices (92701)
a.
(lydrogen Storage - IN 89-44
As requested by NRC internal memorandum dated May 2,1989, the
inspectors reviewed the hydrogen storage facility at Farley to
determine: distance from the hydrogen storage facility to the nearest
safety related structure or air intake, and maximum volume of
gaseous or liquid hydrogen stored on site.
The bulk hydrogen storage facility is located approximately 1300 ft
from the nearest safety related structure which is the diesel
generator building.
The maximum hydrogen stored at this location is
200,000 ft3 in the gaseous form and 1,500 gallons of liquid hydrogen.
The distance between this bulk storage location and safety related
structures should not present a fire or explosion hazard to safety
related plant structures.
A hydrogen cylinder having a maximum capacity of 280 cubic feet ic
stored adjacent to the main steam valve room of each unit.
The
location and arrangement of these cylinders do not present hazards to
safety related components which warrants additional protection.
Additional hydrogen cylinders are installed in a number of plant
locations such as counting rooms, laboratories, etc.
These were
considered in the plant's fire prntection evaluation and found to
pose minimal danger to the plant's safe shutdown capability.
Spare
cylinders for these applications are stored in the warehouse complex
located approximately 700 feet from the nearest safety related
structure.
The maximum quantity of hyarogen in this location is
approximately 2,000 cubic feet and does not appear to present a fire or
exposure hazard to safety related plant structures.
The licensee is evaluating this IN.
b.
Potential Failure of Westinghouse Steam Generator Tube Mechanical
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Plugs - IN 89-33
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Upon receipt of this information notice, the licensee immediately
initiated an investigation to determine if any of the plugs installed
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in the Unit 2 steam generators were of the heat numbers identified by
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Westinghouse as being susceptible to cracks. The following number of
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susceptible plugs were found installed in Unit 2:
S/G
S/G
S/G
Heat No.
A
B
C
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3962
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15
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4523
13
35
61
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Unknown
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103
2
TOTALS
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The above plugs were modified by installation of a Westinghouse
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designed " plug-in-plug" modification, except five heat No. 4523 plugs
in steam generator C which were replaced witn new plugs. These five
plugs were removed and returned to Westinghouse for further
evaluation.
Also, one leaking plug in steam generator 2C of heat
No. 3962 was removed and replaced by a new plug.
The licensee's evaluation of IN 89-33 is in progress to determine the
number of plugs installed in Unit I which require the " plug-in-plug"
,
tradification during the September 1989 refueling outage.
(
c.
Potential LOCA at High and Low Pressure Interfaces from Fire Camage -
Valves 870iA, 87018, 8702A and 87028 in the surtion piping from the
RCS to RHR pumps are located in containment ar.- are maintained in the
closed position, except when the RHR system is in operation.
These
.
valves have interlocks to prevent inadvertent over pressurization of
3
the RHR system, depressurization of RCS or overflow and dilution of
)
the RWST. The RHR inlet isolation valves are designed to automatically
I
close if RCS pressure exceeds 700 psig.
The inspector's were
concerned that the power supply for the above valves at Farley is not
removed when the units are above modes 4 to avoid the type concerns
i
discussed in IN 87-50 involving the potential for " shorts" that could
)
cause an unplanned operation of these interface valves in the event
I
of a fire in certain locations.
The licensee stated that power is
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maintained on these valves to facilitate initiation of the RHR system
)
when required.
The licensee's current evaluation found that this
j
information notice was not applicable to Farley.
The inspectors
suggested that this item be re-evaluated. This concern is identified
i
as Inspector Followup Item 348/89-11-01, 364/89-11-03, Licensee's
Reevaluation of Information Notice 87-50.
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9.
Onsite Followup of Written Reports of Nonroutine Events (92700)
a.
(Closed) 348,364/P2189-01, Slow close lever rebound spring for Brown
Boveri; M'd power distribution K-Line electrically operated K-225
through K-7000 circuit breakers.
The licensee conducted . an
evalut.cion and determine that ABB K-Line circuit breakers were not
installed at Farley.
Therefore, this item is not applicable at
Farley and is closed.
10. Action on Previous Inspection Findings (92702)
a.
(Closed) Deviation 348,364/89-32-01, Receipt of high radiation signal
by radiation monitors in control room ventilation system does not
sound an alarm in the control room.
The licensee evaluated this discrepancy and determined that the alarm
commitment could be removed from the FSAR.
However, further
management review determined that an audible alart would be a
desirable enhancement and this alarm has been installed.
This item
is closed.
b.
(0 pen) Unresolved Item 348,364/87-33-01, Correct fuse size
designation on as built drawings.
NRC Report 348,364/89-05
requested a written response from the licensee indicating the status,
corrective action, and significance of this item.
The licensee's
response of April 24, 1989 indicated that an evaluation of plant
fuses will be made to determine if correct fuse sizes and an
inspection will be made to verify that the correct fuses are
installed. A fuse manual is to be written to assure that the correct
fuses are installed during future maintenance. This fuse manual will
supersede existing design drawings and manuals as the controlling
document for fuses.
The manual is scheduled to be completed and
fully implemented by September 29, 1989.
The fuse inspection and
evaluation program results are scheduled to be completed by
December 31, 1989.
This item will be re-evaluated upon completion
of the licensee's review. This item remains open.
11. Exit Interview
The inspection scope and findings were summarized during management
interviews throughout the report period and on May 10, with the plant
manager and selected members of his staff.
The inspection findings were
discussed in detail.
The licensee acknowledged the inspection findings
and did not identify as proprietary any material reviewed by the
inspectors during this inspection.
Licensee was informed that the items discussed in paragraph Nos. 6, 9 and
10 were closed.
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Item Number
Description and Reference
364/89-11-01
(OPEN) Violation: Failure to maintain
containment integrity - paragraph 3.b.(1).
364/89-11-02
(OPEN) Violation: Failure to follow
procedures while preparing the solid state
protection system for a surveillance test -
paragraph 3.b.(2).
348/89-11-01
(OPEN) Inspector Followup Item: Licensee's
364/89-11-03
re-evaluation of IN 87-50, high-low pressure
interface - paragraph 8.c.
348/89-11-02
(CLOSED) Non Cited Violation:
Failure to
take grab samples when - radiation monitors
were inoperable.
12. Acronyms and Abbreviations
.i
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-
Abnormal Operating Procedure
A0P
-
Administrative Procedure
-
APC0 -
Alabama Power Company
Code of federal Regulations
CFR
-
CCW -
Cor,ponent Cooling Water
-
Design Change
DR
Deviation Report
-
ECP -
Emargency Contingency Procedure
Emergency Plant Implementing Procedure
EIP
-
Environmental Qualifications
-
ESF -
Engineered Safety Features
Engineering Work Request
-
Fahrenheit
F
-
GPM
Gallons Per Minute
-
IN
Information Notice
-
Inservice Inspection
-
Inservice Test
-
LC0 -
Limiting Condition for Operation
MOV -
Motor-0perated Valve
MOVATS - Motor-0perated Valve Actuation Testing
Maintenance Work Request
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MWR
-
Nonconformance Report
1
-
'
Nuclear Regulatory Commission
NRC
-
NRR -
NRC Office of Nuclear Reactor Regulation
Plant Modifications Department
-
Quality Assurance
-
Quality Control
-
RCP -
Radiation Control and Protection Procedure
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RHR -
'
Resistance Temperature Detector
-
,
Safety Injection
!
SI-
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SAER -
Safety Audit and Engineering Review
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S/G
-
SSPS -
Solid State Protection System
Solenoid Operated Valve
S0V
-
Surveillance Test Procedure
-
SW.
--
Technical Specification
TS
-
-
WA
Work Authorization
-
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