ML16154A779
| ML16154A779 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/21/1995 |
| From: | Crlenjak R, Harmon P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML16154A778 | List: |
| References | |
| 50-269-95-03, 50-269-95-3, 50-270-95-03, 50-270-95-3, 50-287-95-03, 50-287-95-3, NUDOCS 9505030140 | |
| Download: ML16154A779 (17) | |
See also: IR 05000226/2003025
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report Nos.:
50-269/95-03, 50-270/95-03 and 50-287/95-03
Licensee:
Duke Power Company
422 South Church Street
Charlotte, NC 28242-0001
Docket Nos.: 50-269, 50-270 and 50-287
License Nos.:
Facility Name: Oconee Units 1, 2 and 3
Inspection Conducted:
February 26 - March 25, 1995
Inspector:
/
. E. Hafmon, Senior R dent/nspector
Date Signed
W. K. Poertner, Resident Inspector
L. A. Ke 1ler, Resident Inspector
P. G.,
phrey, Re ident Inspector
Approved by:
- k
/V.rlenA
Chief
Dte Signed
Reactor Projects Branch 3
SUMMARY
Scope:
This routine, resident inspection was conducted in the areas of
plant operations, surveillance testing, maintenance activities,
onsite engineering and technical assistance, plant support,
inspection of open items, and review of licensee event reports.
Inspections were performed during normal and backshift hours and
on weekends.
Results:
An Unresolved Item was identified in Plant Operations regarding
the misalignment of a valve in the Auxiliary Service Water System,
paragraph 2.d.
Within the area of Engineering, one Violation with 2 examples was
identified that involved calculation errors associated with the
maximum power output limits of the Keowee Hydro Units, paragraph
4.b. Additionally, the need for a Technical Specification change
to clarify the number of High Pressure Injection Pumps required
for reactor modes of operation was identified as an Inspector
Followup Item, paragraph 6.a.
Enclosure 2
9505030140 9504216
QPDR
ADOCK 05 0002 9
2
Plant Support activities of the Nuclear Safety Review Board were
well organized and aggressively pursued resolution of plant
problems, paragraph 5.b. In addition, the transportation and
storage of a spent fuel assemblies canister was well planned and
coordinated, paragraph 5.a.
Enclosure 2
REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- B. Peele, Station Manager
- L. Azzarello, Mechanical Systems Engineer
- E. Burchfield, Regulatory Compliance Manager
- D. Coyle, Systems Engineering Manager
J. Davis, Engineering Manager
T. Coutu, Operations Support Manager
- W. Foster, Safety Assurance Manager
- J. Hampton, Vice President, Oconee Site
D. Hubbard, Superintendent, Instrument and Electrical (I&E)
C. Little, Electrical Systems/Equipment Manager
- J. Smith, Regulatory Compliance
- G. Rothenberger, Operations Superintendent
R. Sweigart, Work Control Superintendent
- L. Wilkie, Safety Review Manager
Other licensee employees contacted included technicians, operators,
mechanics, security force members, and staff engineers.
- Attended exit interview.
2.
Plant Operations (71707)
a.
General
The inspectors reviewed plant operations throughout the reporting
period to verify conformance with regulatory requirements,
Technical Specifications (TS), and administrative controls.
Control room logs, shift turnover records, temporary modification
log, and equipment removal and restoration records were reviewed
routinely. Discussions were conducted with plant operations,
maintenance, chemistry, health physics, instrument & electrical
(I&E), and engineering personnel.
Activities within the control rooms were monitored on an almost
daily basis. Inspections were conducted on day and night shifts,
during weekdays and on weekends. Inspectors attended some shift
changes to evaluate shift turnover performance. Actions observed
were conducted as required by the licensee's Administrative
Procedures. The complement of licensed personnel on each shift
inspected met or exceeded the requirements of TS. Operators were
responsive to plant annunciator alarms and were cognizant of plant
conditions.
Enclosure 2
2
Plant tours were taken throughout the reporting period on a
routine basis. During the plant tours, ongoing activities,
housekeeping, security, equipment status, and radiation control
practices were observed.
b.
Plant Status
All three units operated at or near full power throughout the
inspection period.
c.
Reactor Building Tour While At Power
On February 28, 1995, the inspector accompanied licensee personnel
on a tour of accessible areas in the Unit 3 Reactor Building (RB)
while at 100 percent power. The licensee's purpose for the RB
entry was to identify boron leaks in order to plan the necessary
repair activities for the June 1995 refueling outage. The
inspector noted that ALARA (As Low As Reasonable Achievable)
principles were followed and that the radiation protection
requirements were adequate. The inspector did not identify
unsecured equipment or debris, and concluded that the general
cleanliness of the RB was adequate.
d.
Configuration Control Of Plant Equipment
A non-licensed operator found Unit 2 Condenser Circulating Water
valve 2CCW-110 in the closed position on March 23, 1995. The valve
was in the supply line of the Auxiliary Service Water System. It
is required to be in the open position to provide emergency
feedwater to the 2A Once Through Steam Generator per calculation
OSC-2262 (Tornado Protection Analysis) in the event of the
designated emergency. It had been opened and verified open on
October 29, 1994, per operations checklist OP/2/A/1104/12,
Condenser Circulating Water System, Enclosure 4.19, Valve
Checklist, and 4.20, Valve Checklist Verification.
The operator noted the closed valve during a Job Performance
Measures exercise and generated a Problem Identification Process
Report (PIP 2-095-0350). However, the event was similar to that
found by the licensee on August 15, 1994, when the comparable
valves in Units 1 and 3 were found in the closed position. At
that time, the licensee determined that those valves were not
realigned after completion of hydrostatic tests performed on June
2 and 3, 1995. The licensee stated that realignment was thought
to be accomplished under the feedwater system alignment since most
of the valves involved in the hydrostatic test were in the
feedwater system. As a result, the auxiliary service water valves
were missed in the realignment. The reason for the misalignment
of Unit 2 valve 2CCW-110 has not been determined. As a result, an
Enclosure 2
3
Unresolved Item will be opened, URI 50-270/95-03-03, Valve
Configuration.
Within the areas reviewed, the licensee's actions were determined to be
appropriate with the exception of a misaligned valve.
3.
Maintenance and Surveillance Testing (62703 and 61726)
a.
Maintenance activities were observed and/or reviewed during the
reporting period to verify that work was performed by qualified
personnel and that approved procedures adequately described work
that was not within the skill of the craft. Activities,
procedures and work orders (WO) were examined to verify that
proper authorization and clearance to begin work was given,
cleanliness was maintained, contamination exposure was controlled,
equipment was properly returned to service, and limiting
conditions for operation were met.
Maintenance activities observed or reviewed in whole or in part
are as follows:
(1) Investigate And Repair "B" Sullair Compressor,
WO 95020040 Task 01
The inspector observed troubleshooting and repair in
progress associated with the "B" Sullair Compressor which
had dumped oil on the floor. The equipment, which serves as
a backup air supply for the three units at Oconee, was
classified as non-safety and the work effort was performed
in accordance with Duke Class G criteria. The discharge
check valve was suspected of failing in the shut position,
causing the rotor to rotate in the reverse direction and the
subsequent oil dump.
The work order included the performance of maintenance
procedure MP/O/B/3007/009, Compressors - Sullair Frequent
and Periodic Inspections and Preventative Maintenance. This
procedure presented guidelines for monthly, semi-annual and
annual preventive maintenance that included air filter
inspection, cleaning or replacement, oil addition, oil and
oil filter changeouts, and oil separator element
replacement. The maintenance activity was accomplished in
accordance with the controlling procedure and the inspector
did not identify any deficiencies.
(2) Investigate and Repair Delta Pressure On Unit 2 Feedwater
Valves, WO 95020475
Maintenance activities were reviewed by the inspector to
investigate and evaluate an increased pressure differential
Enclosure 2
4
(72 psid vs 35 psid normal) across the 2B steam generator
feedwater control valve (2FDW-41). The activity was
performed on March 13, 1995, and consisted of an evaluation
of pressure transmitter 2FDW PT0032, which measures
differential pressure across the 2A steam generator
feedwater control valve (2FDW-32). The feedwater control
circuitry utilizes input from the differential pressure
transmitters in both feedwater lines to control the
feedwater to the steam generators.
Operation of the feedwater system was placed in manual
control to allow manipulation of the instrument root valves
at pressure transmitter 2FDW PT0032. The purpose was to
check for leakage at the transmitter equalization valve.
Although no leakage was identified, the effort was
successful in the elimination of suspect equipment that
could be causing the unexplained high pressure drop across
control valve 2FDW-41.
The work effort was performed satisfactorily; however, the
condition causing the high pressure differential across the
valve was not determined. The licensee continues to monitor
and investigate this condition.
(3) Replace/Calibrate Reactor Protection System Feedwater Pump
Discharge'Pressure Switches, WO 95005574
Replacement of the Unit 3 Feedwater Pump Discharge Pressure
and Feedwater Pump Control Oil Pressure Switches was
observed in progress by the inspector. The activity was
necessary because the existing pressure switches were
exhibiting excessive drift and had been placed on an
increased calibration frequency. The cause of the
calibration drift was attributed to problems associated with
the switch diaphragm. The switches being replaced were
Static-O-Ring Models 11013805N and 1101380106N, which had a
316 stainless steel welded wetted diaphragm with a 2 layer
polyamide tertiary diaphragm. The replacement switches were
Static-O-Ring Models 1101401231N and 110140123N, which have
316 stainless steel welded wetted diaphragm and 1 layer of
Teflon coated polyamide tertiary diaphragm. (Note:
The
licensee also plans to change out these switches on Units 1
and 2.)
The inspector questioned use of a work order as opposed to a
design change. The concern was that the replacement switch
models may not be documented and later purchases would not
be for the correct equipment. However, engineering had
documented the change and submitted an Acceptable Substitute
Request (Form NPP-212D) to ensure documentation update.
Enclosure 2
5
The inspector determined the activity to be acceptable and
accomplished in accordance with the requirements specified
in the procedure.
(4) Replace Powdex Conductivity Probes, WO 95006002
Task 01
The inspectors reviewed work activities associated with this
corrective maintenance activity. The conductivity probes
were replaced due to frayed electrical connections. Work
activities reviewed were performed satisfactorily. No
deficiencies were noted.
(5) Infrared Thermography Survey of Condensate Booster Pump C,
WO 94073066 Task 01
The inspectors reviewed work activities associated with this
preventive maintenance (PM) activity. This PM is performed
semi-annually and consists of checking the breaker cubicle
for electrical hot spots. The work activity was
accomplished per procedure MP/0/B/3016/012, Using
Inframetric Model 600 Scanner. The maintenance activity was
accomplished in accordance with the controlling procedure
and the inspectors did not identify any deficiencies.
(6) Calibrate Unit 1 Auxiliary Instrument Air (AIA)
Instrumentation, WO 94083457 Task 01
On March 16, 1995, the inspector observed calibration of
various instruments associated with the 1 AIA air
compressor. All activities observed were satisfactory.
(7) Lubrication PM on 3C Spent Fuel Cooling Pump, WO 95017147
Task 01
The inspector verified that the correct lubricant was used
and that the work was accomplished in accordance with the
licensee's generic pump lubrication procedure
(MP/0/A/1840/040). All activities observed were
satisfactory.
b.
The inspectors observed surveillance activities to ensure they
were conducted with approved procedures and in accordance with
site directives. The inspectors reviewed surveillance
performance, as well as system alignments and restorations. The
inspectors assessed the licensee's disposition of any
discrepancies which were identified during the surveillance.
Enclosure 2
6
Surveillance activities observed or reviewed in whole or in part
are as follows:
(1) Turbine Driven Emergency Feedwater Test, PT/2/A/0600/12
The inspector reviewed testing of the Unit 2 turbine driven
emergency feedwater pump on February 27, 1995. The
equipment alignment for the test was in accordance with the
procedure and test instruments were within their calibration
date allowances. The purpose of the test was to demonstrate
operability of the pump as required per TS Sections 3.4,
4.0.4 and 4.9.
The activity was performed to acceptable standards.
(2) Engineered Safeguards System Logic Subsystem 2 High Pressure
Injection and Reactor Building Isolation Channel 2 On-Line
Test, IP/O/A/0310/13A
The inspectors reviewed testing of the Unit 1 Engineered
Safeguards System conducted on March 24, 1995. The
procedure implements the requirements of TS 4.1.1, Table
4.1-1, Item 14. The TS requires that a logic test be
performed monthly. The inspectors verified that the testing
was accomplished in accordance with the procedure and that
the acceptance criteria was met. No deficiencies were
noted.
(3) Unit 1 Motor Driven Emergency Feedwater (MDEFW) Pump Test,
PT/1/A/0600/13
On March 15, 1995, the inspector observed the operability
test of the 1A MDEFW pump. The inspector verified that the
test was performed in accordance with procedure and that all
parameters met the acceptance criteria.
Within the areas reviewed, licensee activities were satisfactory. No
violations or deviations were identified.
4.
Onsite Engineering (37551)
During the inspection period, the inspectors assessed the effectiveness
of the onsite design and engineering processes by reviewing engineering
evaluations, operability determinations, modification packages and other
areas involving the Engineering Department.
a.
Measured Capacities of the 125vdc Control Batteries
Unit 3 control battery 3CA was tested on March 1, 1995. Its
measured capacity was 77.97 percent. Capacity should be at 80
Enclosure 2
7
percent or greater to comply with design calculations and the
Design Basis Documents. As addressed in Inspection Report 50
269,270,287/95-01, the licensee performed calculation OSC-5938 on
January 5, 1995, to determine acceptable operability of the 2CB
battery when it was tested and found to be at 77 percent capacity.
This calculation utilized the maximum design basis load for all
batteries and was applied to the 3CA battery as well.
In addition
to control batteries 3CA and 2CB, a third battery (2CA) was
measured at 80.2 percent.capacity.
All six control batteries (Exide FTC-23) are utilized as a backup
for various safety-related equipment during an accident involving
a loss of electrical power. Representatives from Exide have been
involved and the licensee reported that Exide will provide
replacements for those battery cells which tested low. However, a
delivery schedule was not available at the end of the inspection
period. Inspection of this issue will continue under Inspector
Followup Item 50-270/95-01-01, Control Battery Capacities.
b.
Inadequate QA-1 Calculations Involving Keowee Hydro Unit (KHU)
Power Limits
On October 12, 1992, the licensee discovered a single failure
vulnerability due to the "zone overlap" of certain differential
current protective relays (LER 269/92-16). As discussed in LER 269/92-16, and paragraph 7.a of this report, the corrective action
to eliminate this vulnerability (opening the breaker disconnects
on the overhead path of the KHU aligned to the underground path)
resulted in the inability to simultaneously generate both KHUs to
the grid. The licensee, for financial reasons, determined that
the inability to simultaneously generate both KHUs to the grid was
unacceptable. Therefore, the development of a modification to
preclude this single failure and allow simultaneous generation to
the grid was initiated. This modification was eventually
designated NSM-52966.
On January 11, 1993, the on-going "Keowee Single Failure Analysis"
identified the technical inoperability of the KHUs while
generating to the grid during certain power/lake level
combinations, due to turbine overspeed (LER 269/93-01). A KHU
generating to the system grid at a high load when an emergency
start initiated, would separate from the grid and overspeed.
Under certain power/lake level combinations the speed would reach
the overspeed trip setpoint, which trips the unit and opens the
field breakers and prevents their reclosure. As a result,
operation of a KHU to the system grid was limited to 66 MW. This
limit was documented in calculation OSC-6003, Revision 0, "Keowee
Operating Limits to Prevent Overspeed Due to Load Rejection."
The
basis for the 66 MW limit was load rejection test data from
initial startup. These initial startup load rejection tests were
Enclosure 2
8
conducted at 22, 44, 66, 88 and 100 MW. Since the test at 88 MW
resulted in an overspeed trip and the test at 66 MW did not, 66 MW
was chosen as the load limit.
On May 20, 1993, OSC-6003 was revised (Revision 1) to change the
66 MW operating limit to 75 MW. This was based on load rejection
tests conducted during April 1993 and calculation KC-UNIT-1-2
0097.
On May 16, 1994, another scenario was found that could render the
emergency power system inoperable due to overspeeding following
separation from the grid (LER 269/93-01, Revision 1).
The
scenario involves both KHUs generating to the grid when a
LOCA/LOOP occurs. Both units receive an emergency start signal,
separate from the grid, and then overspeed due to load rejection.
This results in above normal frequency output such that the
electrical supply to certain safety-related motors yields a lower
starting torque, requiring longer time to accelerate which could
result in a motor trip on overcurrent. The licensee expanded the
scope of NSM-52966 to eliminate the possibility of
overspeed/overfrequency while allowing simultaneous generation of
both KHUs to the grid.
On January 19, 1995, the licensee made a presentation to NRC
regarding their proposed modification (NSM-52966) that, if
implemented, would enable them to generate both KHUs
simultaneously to the grid. At this meeting, members of NRC's
Electrical Engineering Licensing Branch (EELB) expressed concern
over the licensee's 75 MW operating limit in that it appeared to
be non-conservative when compared to the 1971 startup test data.
At the meeting, EELB representatives requested additional
information on benchmarking of the computer model and instrument
uncertainties used in the calculations that established the 75 MW
limit. The EELB representatives agreed to visit Oconee to review
the requested information.
On January 26, 1995, EELB representatives visited the Oconee site
to review the relevant calculations associated with NSM-52966.
The EELB representatives noted that calculation KC-UNIT-1-2-0106,
"Keowee Power Operating Restrictions for NSM-52966" was non
conservative in that it did not properly account for potential
instrument inaccuracies associated with previous test data used in
this calculation. The EELB representatives requested that the
licensee review the other QA-1 calculation associated with
overspeed (OSC-6003) to see if it was also non-conservative. The
licensee subsequently determined that OSC-6003, Revision 1, was
non-conservative in that it did not take into account instrument
accuracy uncertainties for those instruments used to collect test
data, and a non-conservative instrument accuracy was used for the
KHU watt meter. On January 27, 1995, the licensee suspended KHU
Enclosure 2
III9
generation to the grid until the load limit calculation could be
revised. On January 27, 1995, OSC-6003 was revised (Revision 2)
to establish a new load limit of 56 MW. On March 12, 1995, the
licensee determined that the KHUs had been past inoperable for 31
hours and 3 minutes since the 75 MW load limits were imposed in
May 1993. The inspectors were concerned that there were errors in
QA-1 calculations and that these calculations were independently
reviewed and approved without identifying these errors. The use
of non-conservative instrument accuracies in OSC-6003, Revision 1,
is identified as Example 1 of Violation 50-269,270,287/95-03-02:
Calculation Errors Associated with Keowee Output Limit.
OSC-6003, was revised (Revision 3) on March 14, 1995, in order to
raise the maximum power output limit for a KHU from 56 MW to 69
MW. The statistical analysis methodology utilized to reduce
instrument accuracy uncertainties was incorrect for this revision.
Utilizing the appropriate statistical methodology should have
resulted in a maximum power limit of 68 MW. This calculation was
independently reviewed and approved without identifying the error.
The non-conservative limit of 69 MW was subsequently incorporated
into OP/0/A/2000/041, "Keowee-Modes Of Operation" on March 14,
1995. The use of an incorrect statistical method is identified as
Example 2 of Violation 50-269,270,287/95-03-02: Calculation
Errors Associated with Keowee Outp-ut Limit.
Within the areas reviewed, one violation was identified.
5.
Plant Support (71750 and 40500)
a.
Provide Support For Spent Fuel Shipment 3/6 ISFSI, WO 95018043
Task 01
The inspector reviewed the licensee's efforts during the
transportation and storage of a canister containing 24 spent fuel
assemblies on March 17, 1995. The canister had been installed in
a cask for transporting from the Unit 1-2 spent fuel pool to the
onsite Independent Spent Fuel Storage Installation (ISFSI).
Maintenance Procedure MP/0/A/1500/007, Independent Spent Fuel
Storage Installation, was utilized and complied with during the
shipping and storage activity.
Thirty-two canisters have been stored in the dry storage facility
which has capacity to store a total of 40 canisters. However, the
licensee informed the inspector of their plans to construct 20
additional ISFSI modules during 1995.
The activity was conducted in accordance with the procedure and
was well planned and coordinated.
Enclosure 2
b.
Self-Assessment
During the inspection period, the inspectors attended several of
the Nuclear Safety Review Board (NSRB) meetings conducted on-site
March 22 and 23, 1995. The NSRB reviewed a wide range of issues
and questioned the station staff closely regarding each.
The
meeting was well organized and placed appropriate emphasis on
issues and programs important to safety. Board members
aggressively pursued answers to their questions and resolution of
their concerns.
6.
Inspection of Open Items (92901, 92902 and 92904)
The following open items were reviewed using licensee reports,
inspection records, and discussions with licensee personnel, as
appropriate:
a.
(Closed) Unresolved Item (URI) 50-269,270,287/90-30-01:
Clarification of Technical Specification 3.3.1
This item addressed high pressure injection (HPI) operability
requirements below 60 percent full power operation. TS 3.3.1
requires only two HPI pumps be operable below 60 percent power and
that three HPI pumps be operable above 60 percent power. In
November 1990, the licensee identified that below 60 percent power
an injection line nozzle break could result in insufficient flow
to the reactor core assuming a single failure if only two HPI
pumps were operable. The licensee reported this condition per LER 269/90-15 and established administrative controls to require that
three HPI pumps be operable prior to exceeding 350 degrees F to
ensure adequate core cooling flow could be obtained assuming a
single failure. This item was reviewed for enforcement action in
Inspection Report 50-269,270,287/90-34 and a non-cited violation
was issued. This URI remained open pending completion of a TS
change to clarify the high pressure injection system operability
requirements. The licensee had not completed the TS change
submittal to revise TS 3.3.1 as of the end of this inspection
period. Since the enforcement aspects of this URI have been
reviewed previously and a Non-cited Violation was issued, this URI
is closed. The review/submittal of the revision to TS 3.3.1 is
identified as Inspector Followup Item 50-269,270,287/95-03-01:
Clarification of TS 3.3.1.
b.
(Closed) Violation 50-269/93-03-01, Failure to Follow Refueling
Procedure
This violation involved two separate instances of mispositioned
fuel assemblies during the December 1992 refueling outage core
off-load and reload. The licensee determined that both instances
were caused by human error on the part of the refueling bridge
Enclosure 2
11
trolley operators. The operators failed to perform an adequate
self check when positioning the bridge trolley.
The immediate corrective action was to properly place the
mispositioned fuel assemblies. In addition, refueling procedures
(OP/1,2,3,/A 1502/07) were revised to require independent
verification of bridge trolley position by an individual other
than the operator.
After completing the procedure change, the refueling outages for
Units 2 and 3 were completed with no mispositioning events.
However, In May of 1994, during the Unit 1 refueling outage, two
more instances of mispositioning fuel assemblies occurred. This
latest event resulted in Violation 50-269,270,287/94-16-01 and the
imposition of a civil penalty due to the ineffectiveness of
previous corrective actions. As followup on this matter will be
pursued under this latest violation, Violation 50-269/93-03-01 is
considered closed.
c.
(Closed) Violation 50-269,287/93-03-02, Failure to Follow
Procedures, Two Examples
The two examples of this Violation, associated corrective actions,
and inspector reviews are described in the closure statements of
the applicable Licensee Event Reports (LER) in Paragraph 7 below.
Example 1 involved the operators mistakenly leaving the Emergency
Feedwater (EFW) control valves in the manual position (LER 287/93
01).
Example 2 involved the inadvertent actuation of an
Engineered Safety Feature (ESF) actuation signal during a cooldown
and depressurization of the Unit 1 reactor coolant system (LER 269/93-02). Accordingly, Violation 50-269,287/93-03-02 is closed.
7.
Review of Licensee Event Reports (92700)
The below listed LERs were reviewed to determine if the information
provided met NRC requirements. The determination included: adequacy of
description, compliance with TS and regulatory requirements, corrective
actions taken, existence of potential generic problems, reporting
requirements satisfied, and the relative safety significance of each
event. The following LERs were closed:
a.
(Closed) LER 269/92-16:
Postulated Single Failure That Would
Result in the Loss Of Emergency Power System as Result of a Design
Deficiency
On October 12, 1992, during the performance of a single failure
analysis for the Keowee Emergency Power System, the licensee
determined that the potential existed for a single fault to cause
a loss of both Oconee emergency power paths (overhead and
Enclosure 2
12
underground). This vulnerability existed since plant construction
and was the result of a design oversight.
A network of current transformers, differential relays and lockout
relays are employed to monitor and isolate faults on the Keowee
electrical distribution busses.
Faults are detected by comparing
the electrical current balance for various zones within the
electrical distribution system. The licensee's design employs
overlapping of various protective zones to ensure protection of
the entire power system. There are protective zone overlaps
located at the Keowee overhead air circuit breakers (ACBs). The
licensee determined that a fault occurring within the overlap
region of the overhead path breaker (ACB-1 or 2) associated with
the Keowee unit aligned to the underground path, could disable
both the overhead and underground paths. Due to this particular
zone protection overlap, both the generator (87G) and the
transformer (87T) zone protection relays would detect the
postulated fault. The 87T relay would lock out the overhead power
path by opening both KHUs' overhead breakers. The 87G relay would
open (and lock out) both the underground path breaker and the
overhead breaker (also opened by the 87T) for the KHU aligned to
the underground path. This would leave only the underground path
for the Keowee unit originally aligned to the overhead path, which
would require manually closing in the underground path breaker
associated with that unit. No credit is given in the
design/licensing basis for manual manipulations in the LOCA/LOOP
accident.
To preclude this single failure vulnerability, the licensee
established administrative requirements to open the disconnects
for the overhead breaker associated with the unit aligned to the
underground path. The inspectors verified that this configuration
removes the single failure vulnerability in question by removing
the zone overlap region for the unit aligned to the underground
path. Additionally, the inspectors verified that removing this
particular overlap region does not impact the ability of the
protective relaying to protect the remaining zone areas. This
configuration is acceptable for nuclear safety, but prevents the
licensee from generating both KHUs to the commercial grid at the
same time. Since Keowee is routinely used during peak demand
periods, the licensee has proposed a modification to the Keowee
breaker protective relaying circuitry that would eliminate the
single failure vulnerability and allow simultaneous generation of
both KHUs to the grid. This proposed modification (NSM-52966) is
currently being reviewed by NRR.
Another corrective action discussed in this LER was the completion
of the single failure analysis for the Keowee electrical
distribution system. The inspectors verified that the single
failure analysis was completed. The inspectors did not evaluate
Enclosure 2
II
13
the adequacy of this analysis other than the specific concerns
addressed by this LER.
b.
(Closed) LER 269/93-02, Operator Inattention to Detail Results in
an Unplanned Protective System Actuation During Unit Shutdown
On January 29, 1993, Unit 1 operators were performing procedure
OP/1/A/1102/10, Controlling Procedure for Unit Shutdown, to cool
down and depressurize Unit 1's Reactor Coolant System (RCS) to
repair a leaking core flood tank isolation valve. The reactor was
shut down with Group 1 rods at 50 percent withdrawn in accordance
with the procedure. Shortly after he began RCS depressurization
via manual operation of pressurizer spray, the operator's
attention was diverted by a control rod drive (CRD) position
indication problem. RCS pressure decreased below the 1810 psig
Low RCS Pressure Trip setpoint, and a trip signal was generated.
After responding to the unexpected trip, operators continued the
cooldown and depressurization in accordance with the procedure.
As part of the corrective actions, the operator and the Control
Room Supervisor prepared and presented a "Lessons Learned"
briefing to all 5 operating crews. Additionally, the operating
procedure was revised to require inserting the Group 1 control
rods at 1900 psig in the shutdown sequence. The previous revision
had required the operator to stop the depressurization prior to
reaching the trip setpoint, and insert the rods prior to
proceeding. This incident was cited as Example 2 of Violation
50-269,287/93-03-02, Failure to Follow Procedures. The inspector
reviewed the revised procedure and verified that training records
reflected that the shift briefings had been conducted.
c.
(Closed) LER 269/93-03, Design Deficiency Results in the Technical
Inoperability of the Alternate Reactor Coolant Makeup System
On February 18, 1993, the licensee determined that the Secondary
Shutdown Facility (SSF) Reactor Coolant Makeup Pump (RCMU) for
Unit 1 was inoperable due to excessive nitrogen preload pressure
in the suction stabilizer accumulator's bladder. This condition
could have prevented the bladder and the pump from operating
correctly during a postulated Engineered Safety Feature (ESF)
actuation.
The root cause of this event was determined to be an oversight in
the bladder design calculation. The design calculation had
neglected to include the temperature effect on the bladder
pressure when the RCMU pump is aligned to take suction on the
Spent Fuel Pool during certain accident scenarios.
Corrective actions included venting the excess pressure from the
bladder and completing the scheduled Design Basis Documentation
Enclosure 2
II
14
(DBD) study for the SSF and the RCMU system. The inspector
verified that the bladder had been vented to the new setpoint on
February 21, 1993, and that the DBD was completed and issued in
November 1994. This item is closed.
d.
(Closed) LER 287/93-01, Reactor Trip on Lost Signal Due to
Technician Inattention to Detail, Followed by Operator
Misalignment of Automatic Emergency Feedwater Paths
On January 26, 1993, Unit 3 tripped from 100 percent power when a
technician's error during troubleshooting created a false low
turbine generator output (Megawatt) signal.
The false signal
caused the Integrated Control System (ICS) to open the turbine
control valves wide open, increase reactor power, and increase
feedwater demand in an attempt to recover the "lost" generator
megawatts. The ensuing transient resulted in reducing main steam
pressure and a resultant decrease in the steam driven main
feedwater pumps' output pressure. The Accident Mitigation System
Actuation Circuitry (AMSAC) started the Emergency Feedwater (EFW)
pumps and tripped the main turbine generator, which tripped the
reactor. The unit trip response was normal.
Following the trip, control room operators took manual control of
the EFW control valves to prevent overcooling the RCS, but
neglected to replace the control valve selector switch back in
auto as required by procedure. Leaving the selector switch in
manual would have prevented automatic response of the EFW system
if called upon. This failure of the operators to follow
procedures was identified as Example 1 of NRC Violation
50-269,287/93-03-02, Failure to Follow Procedures.
The technicians had been using a multimeter on a power factor
meter transformer in Unit 3. During this troubleshooting effort
the technicians had inadvertently selected the multimeter's
function to read current instead of voltage. This caused the
Transformer's "X" and "Y" phase fuses to blow and indicate a loss
of Generator Megawatt output to the ICS.
Corrective actions included inserting blank plugs in the current
measuring jacks of all controlled multimeters. The inspector
verified that the tool issue room had installed the blank plugs.
This issue is closed.
8.
Exit Interview
The inspection scope and findings were summarized on March 29, 1995,
with those persons indicated in paragraph 1 above. The inspectors
described the areas inspected and discussed in detail the inspection
findings in the Summary and listed below. The licensee did not identify
Enclosure 2
15
as proprietary any of the material provided to or reviewed by the
inspectors during this inspection.
Item Number
Status
Description/Reference Paragraph
Inspector Followup
Open
Clarification of Technical
Item 269,270,287/
Specification 3.3.1. (paragraph
95-03-01
6.a.)
Violation 269,270,287/
Open
Calculation Errors Associated
95-03-02
With Keowee Output Limit, Two
Examples (paragraph 4.b.)
Unresolved Item 270/
Open
Valve Configuration (paragraph 2.d.)
95-03-03
Unresolved Item 269,
Closed
Clarification Of Technical
270,287/90-30-01
Specification 3.3.1. (paragraph
6.a.)
Violation 269/93-03-01
Closed
Failure to Follow Refueling
Procedure (paragraph 6.b.)
Violation, 269,287/
Closed
FaiTure to Follow Procedures, Two
93-03-02
Examples (paragraph 6.c.)
Closed
Postulated Single Failure That Would
Result in the Loss Of Emergency
Power System as Result of a Design
Deficiency (paragraph 7.a.)
Closed
Operator Inattention to Detail
Results in an Unplanned Protective
System Actuation During Unit
Shutdown (paragraph 7.b.)
Closed
Design Deficiency Results in the
Technical Inoperability of the
Alternate Reactor Coolant Makeup
System (paragraph 7.c.)
Closed
Reactor Trip on Lost Signal Due to
Technician Inattention to Detail,
Followed by Operator Misalignment of
Automatic Emergency Feedwater Paths
(paragraph 7.d.)
Inspector Followup
Open
Control Battery Capacities
Item 270/95-01-01
(paragraph 4.a)
Enclosure 2