ML20141J889
| ML20141J889 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/15/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20141J860 | List: |
| References | |
| 50-285-97-15, NUDOCS 9708210387 | |
| Download: ML20141J889 (17) | |
See also: IR 05000285/1997015
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ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No:
50 285
License No: DPR 40
Report No:
50 285/97-15
Licensee:
Omaha Public Power District
Fort Calhoun Station FC-2 4 Adm.
P.O. Box 399, Hwy. 75 - North of Fort Calhoun
Fort Calhoun, Nebraska
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Facility:
Fort Calhoun Station
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Location:
Blair, Nebraska
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Dates:
June 15 through August 2,1997
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Inspectors:
W. Walker, Senior Resident inspector
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V. Gaddy, Resident inspector
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Approved:
W. D. Johnson, Chief, Project Branch B
Attachment: Supplemental Information
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9708210387 970815
ADOCK 05000285
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EXECUTIVE SUMMARY
Fort Calhoun Station
NRC Inspection Report 50-285/97-15
This routine announced inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers a 7-week period of resident inspection.
Operations
The performance of the operations staff during this inspection period was good.
Operators were observed operating the plant in a professional manner
(Section 01.1).
The inspectors noted that the material condition of the auxiliary feedwater arid raw
water systems was good. Minor discrepancies between the piping and
instrumentation diagram and operating instruction were noted during the walkdown
of the auxiliary feedwater system (Section O2.2).
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Verification of two tagouts found all components in their correct position
(Section O2.2).
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Ineffective corrective action resulted in a work request sticker not being removed
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from the control room panel as required by procedure. This was a violation
(Section 02.3).
Maintenance
The inspectors observed multiple maintenance activities during the report period.
Overall, the maintenance and surveillance activities were thorough and performed-
professionally (Section M1.1).
Ineffective corrective action resulted in the lower disc wedge of the boric acid
totalizer bypass valve being broken due to overtorquing. This was a second
example of a violation (Section M8.1).
Enaineerina
Engineering personnel performed a sound technical analysis that resulted in a
decrease in the correlated nuclear detector well temperature (Section E1.1).
The licensee provided adequate justification to support the cable routing deviation
identified by the inspector (Section E8.1).
Plant Sucoort
Due to a procedural misunderstanding, the self-contained breathing apparatus
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regulator flow test records were not documented appropriately. Once pointed out
by the inspectors, the test records were correctly documented (Section R2.1).
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While touring the radiation controlled area, the inspectors identified that locked
doors could not be opened due to improper maintenance on the door locks. This
item will remain unresolved (Section R2.2).
The licensee determined that the criticality monitor in the new fuel receipt area was
not sensitive enough to detect a criticality accident (Section P2.1).
The inspectors identified that licensee personnel did not establish any compensatory
measures prior to blocking the sprinkler system in the diesel generator room. This
was a violation (Section F1.1).
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Report Detaih
Summary of Plant Status
The 'brt Calhoun Station began this inspection period operating at essentially 100 percent
po' .Or and operated at that level throughout the inspection period,
l. Operations
01
Conduct of Operations
01.1 General Comments (71707)
Using inspection Procedure 71707, the inspectors conducted frequent reviews of
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ongoing plant operations, in general, the conduct of operations was professional and
safety conscious; specific events and noteworthy observations are detailed in the
sections below.
O2
Operational Status of Facilities and Equipment
O2.1 Review of Eauipment Taaouts (71707)
The inspectors reviewed the following tagouts:
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Serial Number 97-0607, Replace Circuit Board of 125 VDC Battery Charger 2
Serial Number 97-0715, Isolation of East Raw Water Header
The inspectors found that all tags were on the proper components and components
were in the proper tagged position.
02,2 Enaineered Safetv Feature System Walkdown (71707)
The inspectors used laspection Module 71707 to walk down portions of the auxiliary
feedwater and raw water systems. The systems were walked down using the
following drawings and procedures:
Operating Instruction Ol-AFW-1, Auxiliary Feedwmer Actuation System Normal
Operation, Revision 33
Operating instruction Ol-RW-1, Raw Water Normal Operation, Revision 47
Standing Order S0-0-44, Administrative Controls for the Locking of
Components, Revision 53
Drawing 11405-M-100, Raw Water Flow Diagram
Drawing 11405-M 254, Condensate Flow Diagram
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Drawing 11405 M 253, Steam Generator and Blowdown Flow Diagram
The inspectors noted the material condition of the equipment was good. All supports
and seismic restraints were properly anchored and in good condition. Valves and
breakers were verified to be in the correct positions. However, during the auxiliary
feedwater system walkdown, the inspectors noted a few minor discrepancies
between the operating instruction and the piping and instrumentation diagram.
Specifically, the inspectors noted that Valve FW-1317 (emergency feedwater storage
tank condensate makeup bypass valve) was shown as being locked on the operating
instruction, but was not locked on the piping and instrumentation diagram.
Valves FW 333 (emergency feedwater storage tank level alarm upper isolation valve)
and FW-334 (emergency feedwater storage tank level alarm lower isolation valve)
were shown as closed on the operating instruction, but were locked closed on the
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piping and instrumentation diagram. Valves FW 661 (emergency feedwater storage
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tank condensate makeup inlet valve) and FW-652 (emergency feedwater storage tank
demin makeup water inlet valve) were shown as closed on the operating instruction,
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but were shown as open on the piping and instrumentation diagram. The as-found
position of the auxiliary feedwater valves was consistent with the operating -
instruction.
The inspectors inquired about the discrepancies. Licensee personnel stated that they
relied on the operating instruction to show the current plant valve position. The
licensee initiated Document Change Engineering Change Notice 97-246 to change
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the piping and instrumentation diagram to ensure that they match the operating
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instructions.
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02.3 Failure to Remove Work Reouest Sticker
I Joection Scone (71707)
a.
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The inspectors followed up on inadequate corrective action to ensure work request
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stickers were removed from the control room following maintenance,
b.
Observations and Findinas
On June 17,1997, while walking down the control room panels, the inspectors
noted Work Request Sticker 970291 attached to the motor-driven fire pump. Work
request stickers were used to visually identify items requiring maintenance. The
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work request sticker indicated that the pump had tripped on overcurrent. The
inspectors asked a licensed operator if the pump had been repaired. The licensed
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operator stated that he thought the work had been completed. The inspectors asked
that, if the work had been completed, why was the work request sticker still
attached to the control room panel. The inspectors questioned whether all
operational staff were aware of the status of the equipment.
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The inspectors determined that the pump had been repaired and the paperwork t i
been closed out on June 9,1997. Ttye inspectors reviewed Maintenance Work
- Documents 972000 and 972031. These work documents documented work that-
was performed on the pump. The-inspectors reviewed Standing Order SO-M 101,
" Maintenance Work Control," and noted that Step _5.10.8 directed maintenance
_ personnel to remove. work request stickers following the completion of
postmaintenance testing and/or operability testing.
The inspectors identified _a similar incident in December 1995 that is documented in
NRC Inspection Report 50 285/95 24. This involved the failure to remove a work
request sticker from the control room panel following maintenance on Charging _
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Pump CH 1C. The licensee documented this incident on Condition
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Report 199600119.- The condition report indicated that the work request sticker was
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not removed following maintenance and raised the concern that failing to remove the
sticker following repair could potentially result in ope?ators not knowing the status of
. control room equipment. -The licensee performed a root cause analyc's and initiated
eight action items to prevent recurrence.. All the action items were closed.
Corrective action from the December 1995 incident included revising Standing
Order SO-M-101 to require the shift supervisor to verify work request stickers were
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removed from control room panels prior to closing out a maintenance work
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document. This verification should include seeing the removed sticker attached to
- the maintenance work document. This action item was closed in May 1996.
On February 11,1997,' operations p: : _onnel performed a walkdown of the control
room and identified six work request stickers attached to control room panels whose
work documents were either closed or voided. Earlier corrective action did not
prevent recurrence. -This deficiency was documented on Condition
Report 199700155. Although the root cause analysis was completed in March
1996, several subsequent condition reports have documented f ailures to remove
-deficiency stickers following maintenance.
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The corrective actions implemented by the licensee to ensure inat work request
stickers were promptly removed from the control room panels have not been
effective. Failing to ensure adequate corrective action is a violation of 10 CFR -
Part bO, Appendix B, Criterion XVI (50-285/97015-01).
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Conclusions -
The inspectors identified that the corrective action implemented by the licensee to
ensure work request stickers were removed from the control room following
-maintenance was ineffective.
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11. Maintenance
M1
Conduct of Maintenance
M1.1 f2eneral Comments
a.
inspection Scope (62707)
The inspectors observed all or portions of the following activities:
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Replacement of Raw Water Pump AC-10C
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Leak repair on Steam Generator Blowdown Sample Cooler SL-8
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Replacement of lobe oil circulating pump for Diesel Generator 1
Installation of water curtain in auxiliary building stairway
Numerous maintenance activities on diesel-driven fire pump
b.
Observations and Findinas
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The inspectors found the work performed under these activities to be professional
and thorough. All work observed was performed with the work package present and
in active use. On occasion, the inspectors observed system engineers monitoring
progress, and quality control personnel were present whenever required by the
procedures,
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Conclusions
The maintenance activities observed were conducted in a controlled and professional
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M1.2 Surveillance Activities
a.
insoection Scope (61726)
The inspectors observed all or portions of the following activities:
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OP-ST-VA-0001, Monthly Hydrogen Purge System Test, Revision 6
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OP ST-ESF-0009, Channel "A" Safety injection, Containment Spray and
Reciiculation Actuation Signal Test, Revision 28
IC-ST-RC-0001, Functional Test of Acoustic Flow Monitors, Revision 3
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CH-CP-01 -6784, Calibration of Feedwater Heater Number 6 Hydrazine
Analyzer, Revision 2
b.
Observations and Finding
The inspectors noted that the surveillances were performed in accordance with
procedures. The surveillance procedure was present and in use during the
observations. Communications between personnel performing the tests and the
control room operators were good,
c.
Conclusions
The surveillance activities observed by the inspectors were completed in a controlled
manner and in accordance with procedures.
M8
Miscellaneous Maintenance issues
M8.1 (Closed) Inspection Followup item 50-285/97003-03: Broken Boric Acid Totalizer
Bypass Valve CH-462. Specifically, the lower disc wedge of the valve was broken.
This item remained open pending the licensee's search of vendor manuals to identify
specific torquing and valve wrench requirements for manual valves anct the
completion of a root cause analysis to determine why the valve failed. The vendor
manual for this valve indicated that the valve should tightly seat with less than
60 foot-pounds of torque on the handwheel and that no more than 100 foot pounds-
of torque should be applied to the handwheel. The valve vendor indicated that the
valve would begin to fail when 145 foot-pounds of torque were app"=d, and the
lower wedge of the valve would break into five pieces when 165 foot- pounds of
torque were applied, Valve CH-462 was used to isolate the 2-inch recirculation line
for Boric Acid Tank CH 118. Failure of the valve to isolate caused boric acid pump
flow to indicate low and the broken valve pieces had the potential to cause safety
related equipment failures.
On June 20,1997, the licensee completed the root cause analysis to determine what
esused Valve CH-462 to fail. The licensee concluded: the lower wedge of the valve
f ailed due to overtightening in the closed direction; the failure was most likely caused
by an individual using additional leverage when tightening the valve onto its closed
seat; and that physical evidence indicated that Valve CH-462 was broken due to
excessive force on the handwheel.
Additionally, the licensee identified that there were over 1000 manual valves with
specific torquing limitations in use in various applications. The iicensee determined
that potential existed for other similar valve failures to occur. The root cause
analysis indicated that a more restrictive valve wrench program would be
implemented by September 30,1997. The inspectors asked if any interim guidance
had been provided to operators. The inspectors were informed that training had been
provided to operations personnel to sensitize them to valve overtorquing issues.
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In NRC inspection Report 50 285/95-24 the inspectors addressed an inciuent that
occurred on December 18,1995, in which the yoke bushing of Valve CH-172
(Chaigir g Pump CH 1C inlet isolation valve) failed wsile Charging Pump CH4C was
being isolated during the performance of Surveillance Procedure OP-ST-CH-3003,
" Chemical & Volume Control System Pump / Check Valve inservice Test." A similar
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failure occurred to the yoke bushing for Valve CH-174'(Charging Pump CH 1 A inlet
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isolation valve) on December 22 while isolating Charging Pump CH-1 A during the
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performance of the same procedure. The licensee initially told the inspectors they
suspected the valve failures were due to lack of lubricant being applied to the valve's
stem.
Condition Report 199500441 documented the failures of the yoke bushings. The
condition report discussed three actions items to prevent recurrence. Action item 2,
completed May 17,1996, requested that operations and training recommendations
be provided to enhance operator training in the area of valve manipulation to
minimize the potential for overtightening valves. During the training, which was
provided during Training Rotation 96-02, the importance of proper valve operation
was discussed and operators were told that several recent valve failures were the
result of overtightening. A!so discussed were ways to reduce the occurrence of
valve overtightening. Operators were also encouraged-to provide input to system
engineering or operations management if they had ideas to reduce instances of
overtightening valves.
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Corrective actions in response to Condition Report 199500441 were not effective
and did not prevent Valve CH-462 from being damaged by to oversightening.
Licensee review of the failure of Valve CH-462 concluded that there was a need for a
more restrictive valve wrench program. Failing to assure adequate corrective action
is a second example of a violation of 10 CFR Part 50, Appendix B, Criterion XVI
(50 285/97015 01).
M8.2 (Closed) Unresolved item 50-285/96018-03: missing test documentation. The item
remained open pending the location of documentation to support tests performed on
Main Steam Line Radiation Monitor Isolation Valves HCV-921 and -922. Originally,
the licensee could not produce adequate documentation to show that tests
performed by the vendor on these valves were within the scope of the maintenance
work document. The licensee contacted the vendor and retrieved the test
documentation. The inspectors reviewed the test data and verified that tests
periormed by the vendor were within the scope of the maintenance work document.
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Conduct of Engineering
E1.1 Nuclear Detector-Well Coolina
a.
Jnsoection Scoce (37551)=
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The inspectors follow,sd up on the circumstances surrounding a high temperature =
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-alarm on Nuclear Detector Well TIA-733B.
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Observations and Findinos
On July 20,1997, a high temperature alarm was received on Nuclear Detector
Well TIA 7338. .The temperature was approximately 152aF. This represented the
annulus exit temperature from-the nuclear detector _ cooling system. The detector
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well concrete ternperature was then determined from Figure TDB lll 24 in the
- Technical Data Book. The figure correlated the temperature relationship between the -
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detector well_ temperature and the concrete temperature. Based on the correlationi
-the concrete temperature was approximately 147*F. The Technical Specification
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-limit for concrete temperature was 150*F. If this temperature was exceeded, the
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plant would have to be shut down in accordance with Technical Specification 2.01.
The licensee dete mined the primary cause of the high temperature was high river
- temperature. The river temperature was approximately 82*F.
Normally, the plant operated with one component cooling water pump in operation.
Operatio_ns personnel started a second component cooling water pump to-increase
the flow to the nuclear detector well. The increased flow did not reduce the detector
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well temperature.
' Engineering personnel performed an investigation to determine the best resolution to
- the cooling problem.' Engineering personnel concluded that hardware fixes without
extensive modification would be ineffective in gaining sufficient margin for detector
.weil temperature ar'd that changes to the component cooling water. flow balance
posed an unacceptable _ risk to the reactor coolant pump seals. Engineering personnel
then investigated whether the correlated temperature figure in the Technical Data
Book contained sufficient margin to allow the temperature limits to be adjusted
upwards. During the investigation, engineering personnel used the 10 CFR 50.59
- process to more clearly define _the Technical Specification reference temperature as
the concreto bulk temperature. This resulted in the temperature limits being
increased 4.8"F. Engineering personnel then performed another 10 CFR 50.59
applicability screening to revise the' detector well temperaturc versus concrete
temperature figure in the Technical Data Book.
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With the adjusted temperature limits, the new correlated concrete temperature was
approximately 144*F.
ci Conclusions
The inspectors concluded that the actions taken by ugineering personnel to revise -
the detector well and concrete temperature cocrelation figure were technically sound.
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Miscellaneous Engineering issues (92903)
E8.1- (Closed) Unresolved item 50-285/96018-05: potential cable routing deficiencies.
.This item remained open pending the inspectors' review of an additional analysis
provided by engineering personnel. The anal"tts was intended to show that the
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cable routing deviation noted by the inspectors was acceptable and that the analysis
-justified the deviation. The inspectors reviewed the analysis and determined that the
analysis appeared to justify that the deviation was acceptable.
IV. Plant Support
R2
Status of Radiation Protection and Control Facilities and Equipment
R2.1 Resoiratorv Eouloment (71750)
During this inspection period, the inspectors performed an inspection of the
respiratory equipment maintenance program. The maintenance and inspection
frequency for respiratory equipment was controlled by Procedure RP 507, " Inspection
and Maintenance of Respiratory Equipment." The inspectors noted that the
self-contained breathing apparatus regulator flow test records were not completed for
1996l In response to the inspectors' findings,-the licensee performed an
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investigation and determined that the test had been completed but,- due to a-
procedural misunderstanding, the results had been incorrectly documented by
- radiation protection persMnel, in response, radiation protection personnel were
counseled and the procedure was changed to clearly state what tests were required
_ to be documented. No other anomalies were noted
R2.2 Insrection of Areas Not Easilv Accessible in Auxiliarv Buildino
a.-
Insoection Scoce (71750)
On July 1_1,1997, the inspectors selected several rooms in the radiation controlled
area to inspect. These areas were either designated as contaminated or restricted
high radiation areas. The rooms designated as restricted high radiation were locked,
as was one of the' rooms which was used for storage.
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b.
Observations Ed Findinos
The inspectors observed the radiation protection technician attempt to unlock
Shutdown Heat Exchanger Valve Room 15A. The radiation protection technician
was unable to unlock the door. A review of maintenance history on this door and
other doors in the auxiliary building was conducted by maintenance personnel. This ~
review determined that eight doors which had been serviced utilizing
Procedure GM-RM-FP AO1, " Fire Door Lockset inspection and Repetitive
Maintenance," Revision 5, on July 3,1997, had the tumbler assembled backwards.
This did not affect doors which were normally open in that the doors would still
open. However, any door which was locked with the tumble"t reversed could not be
opened using the keys, without disassembling the locking mecnanism. Maintenance-
personnel identified that the door to the volume control tank room, Room 29, was
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also locked and would not open.
The licensee determined that it took approximately 30 minutes to open the doors
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with the appropriate tools,
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The inspectors questione'd the licensee concerning the impact on operations of these
doors being locked and not easily accessible. The licensee reviewed the components
in these rooms and determined that both rooms contained equipment that was
referenced in the abnormal operating procedures and/or emergency operating
procedures. The licensee concluded that, in an accident condition, not having access
to the equipment in the rooms for approximately 30 minutes would not impact plant
safety. The licensee also performed a reportability evaluation and determined that
the condition was not reportable. This item will remain unresolved pending the
inspectors' review of the assumptions in the reportability evaluation and review of
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the abnormal and emergency operating procedures (50-285/97015-02),
c.
Conclusions
During a tour of the radiation controlled area, the inspectors identified that locked
doors could not be opened because the tumblers had been incorrectly installed. This
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item will remain unresolved pending the inspectors' review of the abnormal and
emergency operating procedures.
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P2.1 Criticality Accident Reauirements
a.
insoection Scone (92904)
The inspectors assessed the licenseo's compliance with 10 CFR 70.24. The
inspectors used guidance entitled " Inspection Plan for Compliance with
10 CFR 70.24, ' Criticality Accident Requirements,' at Operating Nuclear Plants."
The inspectors interviewed several licensee personnel and reviewed the following
documents:
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Updated Safety Analysis Report Section 11.2.3.6, " Area Radiation Monitors";
Updated Safety Analysis Report Section S.5.3.3, "New Fuel Storage";
Condition Report 199600382, " Fuel Handling";
Condition Heport 199600990, " Fuel Handling Equipment Operation";
Condition Report 199700796, "New Fuel Storage Rack";
Operating Instruction OI-FH-1, " Fuel Handling Equipment Operation";
Area Radiation Monitor Alarm Setpoint Validation RP 216;
Maintenance Procedure RE-RI FE-0701, " Receipt of New Fuel";
Operations Technical Specification Required Shift
Surveillance UP ST-SHIFT-0001;
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Surveillance Test IC ST-RM-012 " Calibration of New Fuel Uncrating Entrance
Area Radiation Monitor RM-80";
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Surveillance Test IC-ST-RM-0001, " Quarterly Functional Test of Area Radiation
Monitors";
Operating Procedure OP-12, " Fueling Operations"; and
Technical Data Book, Chemistry TDB-IV.8, " Area Monitoring Setpoints."
b.
Observations and Findinas
The inspectors examined the f acility operating licensee. The licensee had not been
granted an exerytion from 10 CFR 70.24 pursuant to 10 CFR 70.24(d).
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The inspectors reviewed the Updated Safety Analysis Report. The Updated Safety
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Analysis Report provided information regarding area radiation monitors. The
inspectors identified no specific commitments pertaining to criticality accidents.
The inspectors reviewed the accidental criticality monitoring and alarm system. The
system used gamma-sensitive radiation detectors which provided a distinct audible
and visual alarm. The criticality / area radiation monitors and locations monitored were
as follows:
RM-80 - New Fuel Storage Rack / Receipt Area
ft".-87 - Spent Fu91 Pool Area
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The inspectors reviewed the calibration and functional test procedures for the
criticality / area radiation monitors. The monitors generated a warning / alert alarm at
5 mrem per hour for RM 80 and 20 mrem per hour for RM 87.
The radiation monitors for the spent fuel pool and new fuel receipt area were
addressed in Technical Specification, Section 3.1, Table 3-3, item 3.
The licensee
performed a channel check of the monitors on a daily basis and a channel functional
test quarterly. Channel calibrations were performed during each refueling outage.
The inspectors reviewed the licensee's practices regarding evacuations and drills
required by 10 CFR 70.24(a)(3). The licensee had maintained and implemented
emergency procedures for evacuation and drills. The licensee had performed drills in
accordance with the emergency response plan and the emergency plan implementing
procedures.
Control room operators were designated with the responsibility for determining the
cause of criticality alarms. The alarm response procedure directed the control room
operators to contact radiation protection personnel to iavestigate the alarm and enter
Abnormal Operating Procedure AOP-09, "High Radioactivity."
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All personnel with unescorted access receive General Employee Training, which
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teaches the employee to evacuate an area should an area radiation monitor activate
and sound and to report it to radiation protection personnel.
Initially, the licensee informed the inspectors that both criticality monitors were
capable of detecting a criticality accident. Subsequently, the licensee performed
additional calculations and determined that, if a criticality accident occurred in the
new fuel storage area, the actual dose rate at Radiation Monitor RM-80 would be
approximately 1.55E-2 mrem per hour. That would not be sufficiently high to
generate an alarm of Radiation Monitor RM 80. Based on the calculation, the
licensee was considering applying for en exemption from 10 CFR 70.24. A criticality
accident in the spent fuel pool could be detected by Radiation Monitor RM-87. This
item will remain open to allow the inspectors to review the licensee's resolution of
this issue (50 285/97015-03).
c.
Conclusions
The licensee determined that an accidental criticality in the new fuel receipt area
-could not be detected by the area criticality monitor. The licensee was evaluating
whether to pursue an exemption from this requirement.
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F1.1 Fire Protection Imnairrhent Reauirements
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a.
Insnection Scope (71750)
The inspectors performed a review of the compensatory measures performed by the
licensee during painting activities in Diesel Generator Room 2.
b.
Observations and Findinas
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On July 9,1997, the inspectors toured Diesel Generator Room 2. The inspectors
noted that painting' activities were ongoing and a scaffold had been erected to
facilitate painting the ceiling and walls. The scaifold decking was blocking the
sprinkler heads used for mitigating a fire in the diesel generator room.
The inspectors questioned the fire protection system engineer concerning
compensatory measures for the diesel generator room, it was determined that a fire
protection impairment permit should have been obtained and implemented prior to
installing the decking to ensure that appropriate compensatory measures were in
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place.
The inspectors reviewed Standing Order S0-G 58, " Control of Fire Protection System
impairments." Step 5.1.4 requires that a fire impairment permit and appropriate
compensatory measures be in place prior to affecting the operability of the fire
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suppression system.
Technical Specification 5.8.1 requires that written procedures be established,
implemented, and maintained covering activities recommended in Regulatory
Guide 1.33. Failing to obtain the necessary fire protection impairment permit and
failing to establish a compensatory fire watch prior to installing the scaffold decking
is a violation of Technical Specification 5.8.1 (50-285/97015-04),
c.
Conclusions
The inspectors identified a violation of the fire protection procedures that required a
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fire protection impairment permit and compensatory fire protection measures prior to
affecting the operability of the fire suppression system in the diesel generator room.
The fire protection systeni engineer promptly initiated the proper impairment permit
and implementation of an hourly fire watch as a compensatory measure.
VI. Manaaement Meetinas
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Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management
on August 5,1997. The licensee acknowledged the findings presented. During the
course of the exit meeting, one issue concerning criticality monitoring was classified
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as an inspection followup item. _ The licensee informed the inspectors that a request
- for exemption from criticality monitoring, as requ: red under 10 CFR 70.24, was in
. final review and would be submitted to the NRC in the near future. The inspectors
will close the followup item af ter licensee submittal and NRC approval of the
-The inspectors asked the licensee whether any materials examined during the
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inspection should be considered proprietary. No proprietary information was
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identified.
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ATTACHMENT
SUPPLEMENTAL INFORMATION
PARTIAL LIST-OF PERSONS CONTACTED-
G. Bishop, Assistant Plant Manager
J. Chase, Manager, Fort Calhoun Station
H. Faulhaber, Manager, Maintenance -
S.'Gambhir, Division Manager, Production Engineering
- J. Gasper, Manager, Nuclear Projects-
B. Hansher, Supervisor,- Station Licensing
J. Herman, Manager, Outage Management
R. Jaworski,- Manager, Design Engineering, Nuclear
R. Phelps, Manager, Station Engineering
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R. Ridenoure, Supervisor, Operations
B. Shubert, Manager, Chemical Services
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INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering
IP 61726: Surveillance Observations
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IP 62707: Maintenance Observations
IP 71707: Plant Observations -
IP 71750: Plant Support Activities -
IP 92903: Followup - Engineering
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IP 92904: Followup - Maintenance
ITEMS OPENED AND CLOSED
Opened
50-285/97015 01
_VIO= inadequate corrective actions (Sections 02.3 and M8.1)
~ 50 285/97015-02
URI inadequate maintenance on locked doors (Section R2.2)
50-285/97015-03_
- IFI' -criticality monitors (Section P2.1)
-50 285/97015-04
VIO failure to complete fire impairment permit and establish
- compensatory fire watch in diesel generator room
(Section F1-.1)
Cigsgd
50-285/96018-05
URI- potential cable routing deficiencies (Section E8.1)
50-285/96018-03
URI missing test documentation (Section M8.2)
50 285/97003 03
IFl_
broken boric acid totalizer bypass valve (Section M8.1)