ML20126J218
| ML20126J218 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/30/1992 |
| From: | Chamberlain D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20126J193 | List: |
| References | |
| 50-285-92-30, NUDOCS 9301060081 | |
| Download: ML20126J218 (17) | |
See also: IR 05000285/1992030
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APENQIX
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-285/92-30
Operating License:
Docket:
50-285
Licensee: Omaha Public Power District
444 South 16th Street Hall
Omaha, Nebraska 68102-2247
Facility Name:
Fort Calhoun Station
Inspection At:
Blair, Nebraska
1
Inspection Conducted: December 7-11, 1992
Inspectors:
P. Goldberg, Reactor Inspector, Engineering Section, Division
of Reactor Safety
D. Kelley, Reactor inspector, Maintenance Section, Division of
Reactor Safety
P. Wagner, Team Leader, Division of Reactor Safety-
Accompanying Personnel:
B. Pendlebury, Consultant, AECL, Ltd.
Approved:
nhuu1/ OLLMVL
!1lb 91
/ Dwight D.' Chan.berlain, Deputy Director
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hDivisionofReactorSafety
Inspection Summary
Areas Inspected:
A routine inspection of the licensee's actions in response to previous
inspection findings and licensee event report issues.
Results:
The licensee's calculations and evaluations related-to the electrical
distribution system were noteworthy.
The Fire Protection System lateraction Checklist included in Procedure
gel-4 was excellent.
No violations or deviations were identified during the inspection.
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9301060001 921231
ADOCK 05000205
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innary of Inigettign_[_tadiagi:
Inspection followup Item 285/9230-01 - AdeqJacy of the 161kV Offsite
power Supply - was opened in paragraph 2.2.
Violation 9210-01 is discussed in paragraph 3 but remains open.
The following Inspection followup Items were closed in paragraph 2:
9101-01 - 4160 Volt Short Circuit Analysis
9101-02 - Degraded Voltage Analysis
9101-03 - Circuit Bretker/ fuse Coordination
9101-04 - 125 Volt dc Voltage Drop Analysis
9101-05 - Fuse Control
9101-06 - Device Ratings
The following Inspection followup Items were closed in paragraph 4:
8942-01 - Surveillance Test Requirements
9001-01 - Emergency feedwater Storage Tank Level
9001-03 - Post Accident Monitoring Instruments
9021-01 - Wiring interlocks
The following Licensee Event Reports (LERs) were closed in paragraph 5:
89-014 - Electrical Cable Modifications
90-007 - Main Steam and feedwater System Piping
90-016 - Overpressurization of auxiliary feedwater system piping
90-022 - Fire Barriers91-003 - Hechanical Containment Penetration H3
91-004 - Offsite Power low Signal
91-005 - Violation of Technical Specification Requirement
91-007 - Circuit Breaker Coordination of 480 Volt System
92-008 - Relief Valves92-010 - Circuit Breaker Coordination of de System
92-012 - Steam Generator Differential Pressure Trip Setpoints
92-016 - Containment Spray Pump Suction llead
92-017 - Cam followers in SBN Type Switches
92-022 - licater Crain Pump Hotor Electrical Cables
Attachment:
Persons Contacted and Exit Heeting
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DETAILS
1
INTRODUCTION
The NRC performed an inspection of the electrical distribution system
functional capabilities (EDSFI) in February 1991 as documented in NRC
Inspection Report 50-285/91-01.
There were some issues identified during the
EDSF1 which required additional licensee action and followup inspection by the
NRC.
In addition, some issues were reported to the NRC as a result of the
licensee's continued review of the EDSFI findings.
This inspection was
conducted to review the licensee's completed actions for those issues related
to the EDSFI and to review the status of some additional issues that had been
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identified.
2 ELECTRICAL DISTRIBtJTION SYSTEM FOLLOWUP INSPECTION (2515/111)
The inspectors reviewed the status of the issues identified during the EDSFI
as part of this inspection effort.
The review of additional issues related to
the EDSFI are documented later in this report.
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2.1
(Closed) Inspection Followgo item 285/9101-01:
Short Circuit Analysis
During the EDSFI, some questions were raised about the capability of various
components (switchgear and circuit breakers) in the 4160V distribution to
wit 1 stand or interrupt potential short circuit currents. The licensee was
verifying and revising the inputs and assumption that were utilized-in
computing the maximum short circuits that could be generated as part of the
on-going design basis reconstitution effort.
The licensee committed to
complete the analysis by December 31, 1991.
During this inspection, the licensee's Engineering Analysis FC-90-055,
Revision 1, was reviewed. The licensee calculated that the largest attainable
short circuit currents on the 4160V buses would occur when the buses were
connected to the main generator. -The inspectors also noted that the
calculations were revised to consider the effects of electrical cable
temperatures of-25"C verses the 90*C temperatures utilized in the earlier
calculations.
The licensee had also initiated operating instructions (A0P-31)
to caution the operators to control the main generator output voltage within
the analyzed limits.
The inspectors observed that the calculated maximum short circuit currents
were within the interrupting capabilities of the switchgear. The inspectors
noted, however, that there was no margin in the momentary interrupting
capability of the Bus lA4 Switchgear and the calculated maximum current.
The inspectors also reviewed the licensee's actions to resolve concerns raised
with the sizing of the Heater Drain Pump Motor's feeder Cables (see Licensee
Event Report 92-022).
The inspectors reviewed Engineering Change
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Notice 92-0311 which directed the replacement of the No. 2 AWG cables with
No. 4/0 AWG cables.
The inspector agreed that the replacement cables would
resolve the fire safety concern.
2.2
1 Closed) Inspection followuo item 285/9101-02:
Deoraded Voltang
Analysis
The EDSFI noted problems witn the assumptions that had been used in the
licensee's calculation of the minimum voltage that would be available to
various components while autonatically sequencing emergency loads.
The
concerns included loading onto the emergency generators (EDGs) and under
degraded offsite power conditions.
The licensee committed to include the
EDSFI observations in a revised degraded voltage study that was being
conducted n part of their design reconstitution effort.
During this inspection, the licensee's Engineering Analysis FC-92-072,
Revision 0, was reviewed. The licensee had calculated the values of voltage
and frequency that would result from the automatic loading of the EDGs at
discrete intervals.
The alculations were performed using the computer
program "ETAP" and indicated that the EDGs would perform as designed.
The
inspectors verified that the assumed loading times and values corresponded to
the as-built design.
In addition, the inspectors utilized the computer
program "MATHCAD" to perform independent calculations.
The results of the
inspectors' calculations showed good correlation with the values obtained by
the licensee.
The inspectors also reviewed Engineering Analysis FC-90-057, Revision 1.
This
analysis evaluated the voltage available at the 4160V buses during automatic
sequencing of emergency loads.
The inspectors noted that the licensee had
increased the Offsite Power Low Signal (0PLS) relay setpoints to approximately
97 percent of nominal voltage to ensure that adequate voltage would be
available. The revised OPLS setpoints, however, required an offsite power
supply voltage of approximately 166.3kV at the onsite switchyard.
The
inspectors reviewed the voltage profiles at.the onsite switchyard that were
obtained in 1984 and those obtained in 1992.
The inspectors noted a decrease
in the available voltage levels had occurred.
The inspectors questioned
whether adequate voltage would be available to automatically sequence safety
loads during an accident condition. llowever, the present condition meets
regulatory requirements _ for safe shutdown power supply capability and the
onsite emergency power sources are appropriate for accident' condition load
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sequencing. The inspectors discussed this concern with licensee personnel and
were informed that a number of options were-being reviewed. The licensee was
in the process of providing an additional 161kV transmission line to the
onsite switchyard to improve redundancy and power capability.
Licensee
personnel also stated that additional adjustments to the OPLS relay setpoints
was being considered.
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The licensee's actions to ensure an adequate source of offsite power for
automatic sequencing of accident loads will be. reviewed with the Office of NRR
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and will be evaluatec iuring a future inspection.
(Inspection followup Item
285/9230-01)
2.3
1 Closed) Inspection followfo item 285/9101-03:
Circuit Breaker
(gprdination Study
The licensee was performing circuit breaker and fuse coordination studies for
the 480V and 120V ac systems and the 125V de system at the time of the EDSFI
The licensee committed to complete those studies and implement any necessary
corrective actions in response to that inspection.
During this inspection, the licensee's studies and corrective actions were
reviewed.
The inspectors reviewed Engineering Analysis FC-91-004, Revision 1,
and found it to be acceptable. The inspectors noted that the licensee had
replaced the trip units from various motor control center supply circuit
breakers.
The modifications were implemented in accordance with Modification
Request FC-88-013 in order to provide acceptable coordination of the trip
function.
The 120V circuits were modified in accordance with FC-91-020. This
modification installed 30 ampere fuses downstream of the load distribution
circuit breakers in Distribution Panels AE-40A and -400.
The inspectors
verified that the addition of these fuses provided the necessary coordination.
The inspectors also verified that the battery output fuses were replaced in
accordance with FC-91-084 on A)ril 20, 1992.
The replacement fuses provided
the necessary coordination wit 1 the battery distribution system.
The
inspectors also noted that Engineering Analysis FC-91-084 dispositioned the
additional de system coordination concerns that had been raised during earlier
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reviews.
The inspectors found the analysis to contain sound bases for the
conclusions.
2.4
(Closed) 10Haction Followun Item 285/9101-04:
Voltane DroD Analysis
for the 125 Volt de System
The EDSFI concluded that insufficient information was available to provide
assurance that all safety-related, de powered devices would have an adeouate
terminal voltage throughout the battery duty cycle.
The licensee committed to
perform a bounding voltage drop analysis for selected devices powered by the
125V de system.
During this inspection, licensee Calculation FC05827, "125VDC System Voltage
Drop Study," Revision 2, dated December 3, 1992, was reviewed.
The licensee
selected limiting components and determined the minimum operational voltage
levels.
The voltage drop occurring in the electrical wiring and-control
devices (e.g., circuit breaker and relay contacts) were calculated and
subtracted from the assumed battery terminal voltage of 105V dc.
The voltage
was determined to be adequate to ensure proper operation of the devices except
- or three motors related to the emergency diesel generators.
However, the
licensee determined that the potential inoperability of the motors would not
prevent the automatic start of an emergency diesel generator.
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The inspectors also noted that the internal shutdown circuitry of the
inverters had been adjusted to lower the setpoint to 104V de to provide margin
to the calculated minimum terminni voltage of 104.57V dc. The inspectors
found the licensee's calculation and related evaluations to be noteworthy and
considered the quality of the engineering effort and its documentation to be a
program strength.
2.5
IClosed) Inspectign_followuo item 285f9101-05:
Fuse Control proorg
The fitst phase of a three phase fuse control effort was underway at the time
of the EDSFl.
The first phase consisted of sampling 5-10 percent of the
station fuses; the remaining two phases were to verify ap)roximately 25
percent of the facility fuses and obtain design data on tie remaining fuses to
compile a fuse list.
During this inspection the inspectors examined the licensee's progress and
determined that all phases had been co.npleted.
The licensee had also
developed a fuse list (approximately 4000 fuses) that had been published,
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addition, the inspectors noted that the licensee had established provisions to
control fuses in the procedures that controlled maintenance and modification
activities.
2.6
1Clg. sed) Insoection Followuo item 285/9101-06:
Ratino of Control
Circuitry Devices
The EDSF1 questioned the capability of various control devices assembled by
equipment vendors. The team was especially interested in those devices
supplied as part of the emergency diesel generator package.
The licensee was
unable to retrieve the necessary electrical ratings for some of those devices.
Therefore, the itcensee committed to review selected devices to ensure that
the ratings were appropriate for their function.
During this inspection, licensee Engineering Analysis FC-92-049, " Analysts of
the Electrical Ratings of Selected EDG Components," was reviewed.
The
inspectors found the analysis to be complete and comprehensive.
The licensee
utilized manufacturers' data when avellable and performed type testing and
inspections to verify findings. During this inspection, the licensee issued
Revision 1 to the analysis to incorporate the recent 125Vdc system voltage
drop study.
The licensee concluded that the selected coniponents were all
adequately rated to ensure proper operation.
The inspectors determined that the licensee-had used a conservative
methodology in evaluating the operability of the selected components.
3 FOLLOWP ON CORRECTIVE ACTIONS FOR A VIOLATION (92702)
3.1 IQpen) Violation 285/92]O-01:
failure to Maintain Adecuate Procedures
for Containment Sumn level Calibration
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An NRC inspection in April 1992 determined that the narrow range containment
sum) ievel instruments LT-599 and LT-600, had not been calibrated by the
met 1od stated in the Technical Specifications.
In 1990, an NRC inspection had
ideatified the same concern with the calibration of these instruments.
The
calibration procedures were revised in response to that inspection finding to
include the Technical Specification requirement. However, during the 1992
refueling outage, the procedures were revised again and the Technical
Specification requirement that had been added, was deleted.
During the April 1992 inspection, Surveillance Procedures IC-ST-WDL-0001 and
IC-ST-kDL-0002 were revised (Revisions 11 and 8) to once again incorporate
physical measurement of the containment sump level.
The licensee also added a
caution note to the procedures stating that changes should not be made without
Plant Review Committee review of the NRC commitment.
In addition, the
licensee performed the surveillance test to the revised procedures.
The licensee's response to the Notice of Violation, dated June 22, 1992,
contained corrective action steps to insure the violation would not reoccur.
The corrective actions included:
1) performing a root cause analysig 2)
issuing a memorandum to all members of the Plant Review Committee and Nuclear
Safety Review Group emphasizing the need to aerform a thorough review of
procedure revisions to insure compilance wit 1 the Technical Specifications; 3)
revising Procedure N00-QP-3, "10 CFR 50.59 Safety Evaluations," to provide
additional guidance in determining Technical Specification compliance; and 4)-
a review of other float level calibration procedures to determine if revisions
were necessary to clarify the method of calibration required by the Technical
Specifications.
The NRC found these proposed actions to be acceptable by
letter dated July 28, 1992.
During this inspection, the " Root Cause and Generic imp 1tcations Analysis
Report, Containment Sump Level Miscalibration," No. IR-920381 dated June 16,
1992, was reviewed.
The inspectors found the report to be a comprehensive and
detailed analysis of the root cause. Quality Procedure N0D-QP-3, "10 CFR 50.59 Safety Evaluation," Revision 11, was also reviewed.
Section 9.4 of the
procedure was revised to include a statement to ensure that the change
complied with the Technical Specifications.
The inspectors also reviewed
Memorandum FC-0792-92 dated July 17, 1992, that was sent to the Plant Review
Committee and Nuclear Skfety Review Group members. This memorandum discussed
the violation and emphasized the importance of reviewing changes to procedures
for compliance with the Technical Specifications. The inspectors' considered
these corrective action items to be complete.
The last corrective action item included a review other float level
calibration procedures to ensure Technical Specification cyliance.
This
item was being tracked by the licensee as commitment identincation No. 920573
and had a scheduled completion-date of December 31,1992.
This violation
remains open pending verification that this corrective action item has been
-completed.
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4 FOLLOWUP (92701)
4.1
1[lgtedL[nmetion followuo item 285/8942-011 Waiver of Surveillans.g.,
Etniremitt.1
During an inspection of the licensee's surveillance procedures and records,
some instances were noted where surveillance requirements had been waived.
The technical basis for the waivers were not always adequately documented and
the authority for waiving survelliance requirements had been granted to the
system engineers.
The Itcensee acknowledged this apparent weakness and
initiated corrective actions.
During this inspection, the inspectors reviewed the licensee's corrective
actions.
The Itcensee redefined the waiver process as a procedure change and
revised Standing Orders G-23, " Surveillance Test Program," and G-30,
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'Setpoint/ Procedure Changes and Generation," to implement the actions.
The
inspectors found these changes to be acceptable in resolving the concern.
4.2
IClosed) Inspection followuo item 285/9001-01:
Electrical SepardlDB
for the Emeroency feedwater Storace Tank level Transmitters
The licensee identified a problem with the separation distance between the
electrical cables for the two level transmitters.
The separation was
acceptable for normal service applications, but did not appear to satisfy the
requirements for redundant applications of Post Accident Monitoring
Instruments specified in Regulatory Guide 1.97.
The licensee performed walkdowns of the instrumentation loops during the 1992
refueling outage and determined that problems existed which required rerouting
of some cables.
The licensee initiated Modification Request FC-92 0ll to
accomplish the necessary rework. The inspectors found the proposed
modifications to be acceptable and verified that the modification request was
scheduled to be completed during the 1993 refueling outage.
4.3
(Closedi
Iqioection followuo item 285/9001-03:
Labelino of Reaulatory
Guide 1.97 Instrumentation
During the inspection of the Post Accident Monitoring Instrumentation, the
inspectors noted that the licensee had not provided the required unique-
identification for the instruments,
The licensee included a provision for
identifying the specified instrum:nts in Modification Request FC-88-22,
" Detailed Control Room Design Review Labeling, Demarcations Himics and Color
Padding Project."
During this inspection, the modifications were verified to have been
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completed. The inspectors also toured the control room simulator and verified
thht the unique orange dot designator had been attached to selected instrument
labels,
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4.4
(Closed) Inspection Followuo item 285/9021-01:
Wirina Interlock on
Diesel caenerator Outout Breaker
During the followup inspection of the loss of shutdown cooling event that
occurred on February 76, 1990, a wiring deficiency in the emergency diesel
gener kor auto-close circuit was identified. The wiring deficiency prevented
the dier.cl generator outoui breaker from closing on an automatic start unless
largo loads, including the ;ow pressure safety injection pump, had tripped.
However, the Irw pressure safety injection pu.np does not load shed when
manually startsd,
Therefore, ne diesel generator output breaker would not
automatically close dar N c loss of offsite power ir.cident if the pump had
been manually started.
During this inspection, the licensee 5s corrective actions were reviewed.
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licensee had implemented Modification Request FC 90-024 to jumper the low
pressure safety injection pump permissive contacts in the diesel generator
output breaker closing circuit. The licensee's evaluations verified that
innediate loading of the low pressure safety injection pump on the diesel
generator bes would have minimal effect on plant response to a loss of offsite
power incic M .
5 ONSITE REVIEW OF LICENSEE EVENT REPORTS (92700)
5.1
(Closed) Licensee Event Report 285/89-014:
Auxiliary Feedwater Wirina
Outside Desian Basis
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During the design basis reconstitution effort, the licensee discovered that
wiring for the wide range steam generator pressure instrument on the auxiliary
feedwater control panel had been routed through the control room,
A fire in
the control room could, therefore, have rendered this instrument inoperable.
In addition, the pressurizer pressure and steam generator narrow range level
indicators were powered from the 'C" inverter that was fed from Battery 1.
However, the use of Battery 1 had not been analyzed in the fire safe shutdown.
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In the interim, the licensee performed a Safety Analysis for Operability (SAO)
and revised the abnormal operating procedures to address forced evacuation of
the control room.- Finsi resolution consisted of a plant modification to
supply analyzed power to the above instruments and rerouting the wiring so it
would not be affected by a control room fire.
During this inspection, the
modifications were verified to have been completed.
5.2
(Closed) Licensee Event Report 285/90-007: Main Feedwater and Main
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The licensee reported concerns
'th the main feedwater and main steam piping
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and supports outside of conta..went that could be overloaded during a seismic
event. An evaluation was performed which determined that the main feedwater
piping and supports were outside the design basis specified in-licensing
correspondence and in the Updated Safety Analysis Report (USAR) for high
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energy line break locations,
in addition, tne main steam piping stress and
support allowable values were outside the USAR design basis.
The licensee implemented short-term corrective actions during the 1990
refueling outage.
These corrective actions included the preparation of SA0
90-04, dated April 9, 1990, which established an interim acceptance criteria
for plant operation.
The inspectors reviewed the SA0 and found it to be
comprehensive in scope.
In addition, two modifications were completed on main
steas and main feedwater supports during the outage.
The inspectors reviewed
Modihcation Requests FC-90-19. Revision 0, " Main Steam Supports in Room 81
and Turbine Building," and FC-89-45, Revisien 0, "Feedwater Supports in
Room 81 ana Turbine Building." The inspectors noted that the modifications
were well written and well documented.
The licensee's long-term corrective actions included:
1) the clarification of
the USAR concerning seismic requirements for the main steam and main feedwater
piping outside of containmer.t. 2) the modification of piping supports to
comply with the design basis, and 3) the initiation of a review of seismic
classification of the systems identified ie the NRC safety evaluation
concerning the potential for flooding caused by ruptures of non-seismic
Class I systems.
The inspectors also reviewed Modification Request FC-90-038,
Revision 1, ' Main Steam and Feedwater Supports Room 81 and Turbine Building.'
The purpose of the modification was to restore the main feedwater and main
steam systems between the isolation valves and the turbine builaing to Seismic
Class I requirements. The nonsafety-related piping and supports were
reanalyzed to design basis requirements and supports were corrected as needed.
In addition, Engineering Analysis FC-92-030 was initiated to review the
seismic classification of systems with the potential for flooding.
This
analysis had not been completed but was being tracked in the licensee's
commitment tracking system as C10 900217/09.
5.3
(Closed) Licensee Event Report 285/90-016:
Potential for
QyJroressurization of Auxiliary feedwater Floina
The licensee reported concerns with the potential for the turbine driven
auxiliary feedwater pump (FW-10) to overspeed and cause the discharge piping
to be overpressurized.
The report stated that the pump had a speed limiter
but did not have an overspeed trip; therefore, a single failure of the speed
limiter could cause the pump to overspeed.
The inspectors reviewed SA0 90-012 dated May 13, 1990.
The SAO justified the
continued plant operation based on Engineering Analysis FC-90-28, Revision 1
"Effect of_ Single Failure of FW-10 Speed Limiting Governor.' The inspectors
noted that the engineering analysis determined that the maximum discharge
pressure would not cause a failure in the auxiliary feedwater piping.
The SA0
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concluded that since the overspeed and overpressure postulated event would not
cause a loss of safety function of the auxiliary feedwater system, no
immediate compensatory measures were required.
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The long-term corrective actions included incorporation of lessons learned
from the event into the 10 CFR Part 50.59 safety evaluations training program
and a permanent resolution to the concern for overpressurization of the:
auxiliary feedwater piping from overspeeding of the pump.
The inspectors
reviewed the Student Handbook Lesson Plan: 2327-07, Revision 3. "10 CFR 50.59
Safety Evaluations,"
and noted that the lessons learned were incorporated
into the lesson plan that satisfied the first comitment.
The licensee
determined that the current configuration of the auxiliary feedwater system
was acceptable and no modification was required.
The determination was
documented in OPPD Position Paper, " Overpressure of Auxiliary Feedwater
Components due to FW-10 Overspeed," dated June 24, 1991. This position was
based on a review of codes concerning the overpressure of piping. A 10
percent increase over design pressure was allowed under conditions of safety
relief valve actuation.
The maximum overpressure expected if pump FW-10 were
to overspeed was less than 5 percent for the most limiting components.
5.4
(Closed) Licensee Event Report 285/90-022:
The licensee reported that some fire barrier penetrations were declared non-
functional due to either a lack of documentation to verify critical parameters
or inconsistencies between the as-built configuration and the approved
configuration. The licensee committed to:
1) perform the necessary
evaluations, 2) repair or replace barriers as necessary, 3) revise affected.
drawings as part of any modifications, and 4) upgrade the procedures related
to fire barrier configuration control.
The inspectors verified that the engineering evaluations utilized the
guidelines of Generic Letter 86-10. " Implementation of Fire Protection
Requirements," and that the modifications had been completed in accordance
with Modification Request FC-90-072, " Repair / Replacement-of Fire Barrier
Seals and Dampers." The inspectors also reviewed Procedures MEl-20. " Fire
Barrier Evaluations," Revision 0; MEl-21. " Fire Barrier Configuration Control,
Revision 0; and gel-4, " Fire Protection Systems Interaction " Revision 0.
The
inspectors determined that the procedures provided detailed guidance to
accomplish the required tasks.
In particular, the inspectors found the Fire
Protection System Interaction Review Questions Checklist included in gel-4 to
be excellent.
5.5
(Closed) Licensee Event Report 285/91-003:
Mechanical Containment
Penetration M-3_putside Desian Basis
The licensee determined that Containment Penetration M-3 (Chemical Volume
Control System) was outside the basis that allowed exclusion from the Type C
leak testing required by 10 CFR 50, A>pendix J.
The basis for the Ty)e C
testing exemption had been based on tie charging line pressure being ligher
than containment pressure during accident conditions.
The most recent
reanalysis of the post-accident containment pressure response and the .
Emergency Operating Procedure (EOP) charging pump criteria indicated that the
previous basis for an exclusion was not valid.
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The licensee completed SA0 91-01 in February 1991 to provide the basis for
continued operation. The E0Ps were revised to require closing the charging
pump manual discharge valves whenever the charging pumps were secured.
In
addition SA0 91-01 and the USAR were revised to include the updated
containment pressure analysis and the design basis for Containment
Penetration M-3.
The licensee had requested exclusion from the Type C l u k testing required by
Appendix J based upon a new justification.
That request had not been approved
at the time of this inspection.
However, the inspectors determined that the
status of the test exclusion request had no bearing on the continued safe
operation of the facility.
5.6
(Closed) Licensee Event Report 285/91-004: Offsite Power low Sianal
Outside Desion Basis
A licensee engineering analysis revealed that the voltage supplied to some
480V safeguards equipment could fall below 87.5 3ercent without the offsite
power low signal (OLPS) being actuated.
Since t11s voltage reduction was
lower than the 90 percent recommended voltage for some safeguards loads, the
licensee determined the plant was outside its design basis.
The licensee's corrective actions included administrative controls of
equipment configurations and bus loadings, as well as raising the OPLS
setpoints. The long-term corrective action was to alter the existing logic
circuitry such that, upon receipt of a safety injection actuation signal,
large equipment not required for accident mitigation would be automatically
load shed.
During this inspection, Modification Request FC-91-008, which changed the load
shed logic, was verified to have been fully implemented and tested during the-
1992 refueling outage. The completion of the modification negated the need
for continued administrative control of equipment configuration and bus
loading.
5.7
(Closed) Licensee Event Report 235/91-005: Violation of Instrument and
Control Technical Specification 2.15
On February 28, 1991, the Instrument and Control technicians were performing
maintenance on Steam Generator Instrument B/PIC-905.
The instrument was
declared inoperable at 25 minutes past midnight- and the trip units were placed
in bypass. At 11:44 a.m., a mechanical jumper (bypass) was installed on the
instrument. At that time it was realized that the bypass should have-been
installed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the instrument being declared inoperable.. The
cause was determined to be wrong interpretation of the requirement by
operations personnel.
The licensee's corrective actions. included training on this event and the
documentation of a-Technical Specification Interpretation. During
this inspection, all corre.ctive actions were verified to have been completed.
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5.8
10hird) Licentes_ty.ettLRengrL2M131-007: Circuit Breaker cosrdination
2L.iD0V_ Distribution Svitsa
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As discussed in paragraph 3.4.4 of the EDSil Report, the licensee identified
and reported pro)1 ems with the overcurrent trip coordination of various 480V
circuit breakers.
The licensee's report provided a commitment to complete a
coordination study and resolve any problems.
The inspectors reviewed the circuit breaker coordination study and the related
modifications as part of Inspection followup Item 285/9101-03 and found them
to be acceptable.
(See paragraph 2.3.)
5.9 (Closed) Licenne Event Report 205/92-008:
Safety Religf Valve ScLooints
Greater lhan Ogalified System Desian Pressure
The licensee reported concerns with four, safety injection system, relief
valves.
Valves SI-187, S1-309, SI-310, and S1-311 had setpoints that were
greater than the system design pressure as qualified by the original
.
hydrostatic tests.
Reviews of relief valve setpoints in the safety injection,
chemical and volume control, raw water, component cooling water, auxiliary
feedwater, and the reactor coolant systems were performed in accordance with
Engineering Assistance Request (EAR)91-097 to determine if any other relief
valves had nonconforming setpoints. No other nonconformances were identified.
During this inspection, the licensee's response to EAR 91-097 for the four
valves identified above was reviewed.
Valves SI-187 and S1-310 had a setpoint
of 600 psig while the piping was qualified to 500 psig by hydrostatic test.
Valve $1-309 had a setpoint of 350 psig while the pipin9 was qualified to
300 psig.
The licensee determined that the piping systems were operable
because the pressure ratings for the most limiting. components exceeded the
setpoints of the relief valves.
Valve SI-311 had a setpoint of 150 psig while
the piping was qualified to 66 psig during the original hydrostatic test.
However, in 1983 a hydrostatic test had been cor, ducted at 190 psig.
The
licensee considered the 1983 test sufficient to qualify the piping for
150 psig.
Modification Request FC-92-009, Revision 0, was prepared to reduce the
setpoints of SI-187 and S1-310 from 600 to 500 psig and SI-309 from 350 to
300 psig.
The inspectors-reviewed the Maintenance Work Orders 910725, 910733,
and 910732, and determined that the setpoints had been changed and that the
valves returned to service during April 1992.-
The licensee's long-term corrective actions included revising plant drawings,
updating design basis documents, and performing the required 10-year
hydrostatic test to verify the acceptability of the piping associated with
Relief Valve S1-311.
The inspectors reviewed the controlled plant drawings of
the relief valves and determined that _the valve setpoints-had been )roperly
revised. The inspectors also reviewed marked-up revisions to the $1utdown
Cooling System and Low Pressure Safety injection System design basis documents
and determined that they had been updated to reflect'the changes in relief
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valve setpoints and component design pressures.
The 10-year hydrostatic test
for the piping associated with 51-311 was completed on September 29, 1992, in
accordance with Surveillance Test Procedure S5-51-51-3002, Revision 4 '(SI)
Class 150 Piping 10 Year Hydrostatic Test." The hydrostatic test pressure was
195 psig which qualified the piping for a design pressure of 150 psig.
5.10 LClgiesD Licenite_ItnaLEspatLZM/22-010:
Circuit _ Breaker and fuse
2pordination of 125Vdc Distribution System
The licensee reported that problems had been identified during the design
basis reconstitution effort with the coordination of 125Vdc circuit breakers
and fuses.
The problems occurred because system coordination had not been
properly considered during the 1980 replacetient of the batteries and their
output fuses or during the 1985 changes to the system.
The licensee replaced the battery output fuses during the 1992 refueling
outage in accordance with Hodification Request FC-91-026.
The replacement
fuses had a slower response time and provided proper system coordination.
The insaectors reviewed the licensee's coordination study and found it to be
accept ule.
(See paragraph 2.3.)
5.11 10101tdLLkensee Event Report 2R5/92-012:
Nonconserative 511am
Etatra. tor Differential Pre 11ure Trin SetDoints
The licensee determined that the setpoints for all four channels of the steam
generator transient protection trip function were greater than that allowed by
the Technical Specifications. The cause was determined to be an inadequate
program for implementing and controlling setpoints.
The licensee's corrective actions included the initiation of a new procedure,
SEl-9, "Setpoint/ Tolerance Change and Review." Additionally, the procedures
for design change, electrical system interaction and configuration control
were revised to stress the areas of calculation input and uncertainty.
During this inspection, all of the corrective actions were verified to have
been implemented.
285191-Qlr _ Insufficient Containment
5.12 LClqirdLLicensee Event Report
a
EDLav Pumo Net positive Suction Head
The licensee reported that the containment spray pumps might not have an
adequate suction head during the recirculation phase of operation. This issue
was also discussed in NRC Inspection Report 50-285/92-09, dated May 15, 1992.
The report documented that the licensee had approved SA0 92-02, " inadequate
Containment Spray Pump Net Positive Suction Head."
The SA0 stated that the
existing licensing basis did not allow consideration of subcooling in net
positive suction head calculations.
However, there would be sufficient net
positive suction head available under all accident conditions if subcooling
were considered in the calculation.
The report also stated that the licensee
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had performed a 10 CFR part 50.59 safety evaluation which determined that no
unreviewed safety question existed.
The licensee's long-term corrective actions consisted of:
1) the mechanical
design engineering group performing a review of the condition and its root
cause analysis to emphasize the importance of confirming USAR assumptions when
performing calculations, and 2) a revision to USAR Section 6.2.1 which would
take credit for liquid subcooling in the net positive suction head
calculation.
The inspectors reviewed Memorandum PED-fC-92-1908, dated June 5,
1992, which stated that mechanical design engineers had reviewed the condition
and the root cause analysis.
In addition, the inspectors reviewed the program
document review signoff sheet with the signatures of the engineers that had
reviewed the package.
The inspectors also reviewed the USAR Section 6.2
revision which took credit for 25 percent of the net positive suction head
available from sump subcooling.
(The previous revision to the USAR did not
take credit for subcooling.) This revision to the USAR was submitted to the
NRC on September 18, 1992.
5.13 IClosed) Lictasee Event Report ZMR2-017: C ackina of Cam followers iD
SMLlype Switches
The licensee discovered an inoperable General Electric Ty>e SBN Switch
associated with a 4160V circuit breaker.
Inspection of tie switch revealed
that the Lexon cam follower was broken rendering the switch inoperable.
Previous industry experience, including NRC Information Notice 80-13, had
addressed cracking of polycarbonate cam followers due to exposure to
hydrocarbons during manufacture. A General Electric service inforr.ation
letter was issued in 1976 reconsnended the replacement of switches that had
been manufactured between 1972 and 1976.
The licensee determined that none of
the installed switches were manufactured during that period.
In response to Information Notice 80-13, the licensee performed an inspection
of one switch and found 50 percent of the cam followers exhibited the cracking
described in the information notice. A detailed inspection program using
fiber optic techniques was initiated, but due to visual clarity problems, the
inspections were abandoned in 1984. During the 1985 refueling outage, 30 of
the 90 of the safety-related switches in the control room were replaced.
Some
cracking was discovered, but it was attributed to stress rather than
hydrocarbon exposure.
No switch failures were discovered and the replacement
of s.vitches was suspended.
The licensee's corrective actions related to this recent switch failure
included the inspection and replacement of switches in the 4160V switchgear,
contcol board and auxiliary panel that were cracked or deteriorated.
Additional inspection and replacement activities were also >cheduled for
implementation during future refueling outages.
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5.14 1 Closed) LiceDsee Event Report 285/92 022:
Heater Drain Pumo Cables
During the electrical system short circuit studies, the licensee discovered
and reported a problem with the sizing of sone electrical cables.
The
licensee determined that a fire could result in a short circuit between the
,
three phases of the cables that supplied power to some components, including
,
the heater drain pumps' motors.
A ' bolted short connection" caused by the
fire could cause the cable jacket temperature to increase above allowable
limits in other areas where the cables were routed.
The high cable jacket
tem)eratures could, in turn, cause problems with other, adjacent, cables.
Alt 1ough the heater drain pump motor cables were not safety-related, they
could impact the operability of adjacent safety-related cables.
Therefore,
the licensee replaced the heater drain pump cables in accordance with
Engineering Change Notice 92-311.
The inspectors reviewed the licensee's analysis and actions as part of
Inspection followup Item 9101-01 (see paragraph 2.1) and found that they
resolved the problem.
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AUAGEUil
1 PERSONS CONTACTED
1.1 Omaha Public Power District Personnel
- R. Andrews, Division Manager, Nuclear Services
- G. Cook, Supervisor, Station Licensing
- S. Gambhir, Division Manager, Production Engineering
- J. Gasper, Manager, Training
- C. Guliani, Supervisor, Dperations Training
- R. Jaworski, Manager, Station Engineering
- L. Kusek, Manager, Nuclear Safety Review
R. Lewis, Principle Engineer, Design Engineering
- D. Li3py, Licensing Engineer
- R. Meiaffey, Principal Engineer, Electrical Engineering
- S. Miller, System Engineer
- R. Hueller, Supervisor, Electrical Engir.eering
- J. O'Connor, Manager, Design Engineering - Electrical
,
- W. Orr, Manager Quality Assurance and Quality Control
- T. Patterson, Manager, fort Calhoun Station
- R. Phelps, Manager, Design Engineering
- R. Short, Manager, Nuclear Licensing and Industry Affairs
- C. Simmons Station Licensing Engineer
- J. Tills, Assistant Manager, fort C41hodn Station
>
1.2 NRC Reaion IV Personnel
- J. Whittemore, Reactor inspector
- Denotes personnel that attended the public exit meeting conducted on
December 11, 1992.
2 EXIT MEETING
An exit meeting was conducted on December 11, 1992.
During this meeting, the
lead-inspector reviewed the scope and findings of-the inspection.
The
licensee did not identify as proprietary any of the materials provided to, or
reviewed by, the inspectors during this inspection.
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