ML18010B156
| ML18010B156 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/02/1993 |
| From: | Christensen H, Darrell Roberts, Tedrow J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18010B153 | List: |
| References | |
| 50-400-93-12, NUDOCS 9308050063 | |
| Download: ML18010B156 (24) | |
See also: IR 05000400/1993012
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W., SUITE 2900
ATLANTA,GEORGIA 303234199
++*++
Report No.;
50-400/93-12
Licensee:
Carolina
Power and Light Company
P. 0.
Box 1551
Raleigh,
NC 27602
Docket No.:
50-400
Facility Name:
Harris
1
Inspection
Conducted:
Hay 15 - June
18,
1993
Inspectors:
J.
Te row, Senior
Resi
t
pector
D. Roberts,
esident
Ins
ctor
Approved by:
H.
ristensen,
C ief
Reactor Projects Section
lA
Division of Reactor
Projects
Licensee
No.:
D
e
igne
v i//W
Date Signed
7
z.
Date Signe
SUMMARY
Scope:
This routine inspection
was conducted
by two resident
inspectors
in the areas
of plant operations,
radiological controls, security, fire protection,
review
of nonconformance
reports,
surveillance observation,
maintenance
observation,
safety
system walkdown, followup of onsite events,
outage activities,
adverse
weather operations,
licensee
event reports,
and licensee
action
on previous
inspection
items.
Numerous facility tours were conducted
and facility
operations
observed.
Some of these tours
and observations
were conducted
on
backshifts.
Results:
Three violations were identified:
Failure to maintain
an electrical
containment penetration
conductor overcurrent protection
device operable,
paragraph 2.c.(1);
Failure to properly implement plant procedures for the
control of locked valves
and for the oper ation of electrical
switchgear,
paragraphs
2.c.(3)
and 6;
and Failure to verify design
accuracy of auxiliary
feedwater recirculation piping, paragraph
10.
Weaknesses
were noted in the areas of scaffolding control, paragraph
2.b.(7)
and in the documentation of a boric acid leakage
inspection,
paragraph
7.
9308050063
930702
ADOCK 05000400
8
2
More problems
were experienced
with 480 volt Asea
Brown Boveri
LK Circuit
breakers,
paragraph 2.c.(2).
Good pre-planning
and
use of equipment
mockups
were evident for containment
work, paragraph
7.
In addition, the simulator was utilized to provide plant
startup/shutdown
training and potential transients.
Also, simulator training
was provided for the conduct of an infrequent turbine test,
paragraph
3.
REPORT DETAILS
Persons
Contacted
Licensee
Employees
D. Batton,
Manager,
Work Control
J. Collins, Manager, Training
- C. Gibson,
Manager,
Programs
and Procedures
H. Hamby, Manager,
Regulatory Compliance
D. HcCarthy,
Manager,
Regulatory Affairs
T. Morton, Manager,
Maintenance
- J. Hoyer, Manager, Site Assessment
J; Nevill, Manager,
Projects
- W. Robinson,
General
Manager,
Harris Plant
- W. Seyler,
Hanager,
Outages
and Modifications
H. Smith,
Manager,
Radwaste
Operation
- D. Tibbitts, Manager,
Operations
G. Vaughn, Vice President,
Harris Nuclear Project
- B. White, Manager,
Environmental
and Radiation Control
W. Wilson, Manager,
Spent Nuclear Fuel
- L. Woods,
Manager,
Technical
Support
- A. Worth, Principal
Engineer,
Nuclear Engineering
NRC Representatives
- J. Johnson,
Deputy Director, Division of Reactor Projects
Other licensee
employees
contacted
included office, operations,
engineering,
maintenance,
chemistry/radiation
and corporate
personnel.
"Attended exit interview
and initialisms used throughout .this report are listed in the
last paragraph.
Review of Plant Operations
(71707)
The plant began this inspection period in power operation
(Mode 1).
At
12:07 a.m.,
on Hay 22,
1993,
a plant shutdown
was performed to repair
blowdown valve 1BD-8 located inside containment.
The
plant was placed in the hot standby condition
(Hode 3) at 12:44 a.m. to
effect repairs.
Following completion of these repairs,
a reactor
startup
was
commenced
and criticality achieved at 3:19 a.m.,
on Hay 22.
Power operation
was resumed
at 1:30 p.m.,
on Hay 23.
The plant
continued in power operation for the remainder of this inspection
period.
'a ~
Shift Logs and Facility Records
/
The inspector reviewed records
and discussed
various entries with
operations
personnel
to verify compliance with the Technical
Specifications
(TS)
and the licensee's
administrative procedures.
2
The following records
were reviewed:
shift supervisor's
log;
control operator's
log; night order book; equipment
record; active clearance
log; grounding device log; temporary
modification log; chemistry daily reports; shift turnover
checklist;
and selected
radwaste
logs.
In addition, the inspector
independently verified clearance
order tagouts.
The inspectors
found the logs to be readable,
well organized,
and
provided sufficient information on .plant status
and events.
Clearance
tagouts
were found to be properly implemented.
During a
review of the shift supervisor's
log, the inspector noted
an entry
for June
7,
1993,
where the shift supervisor
had identified an
expired working copy of procedure
OP-. 126
Extraction
Steam,
and
Steam
Dump Systems)
during
a tour of the steam tunnel
area.
Selected
controlled copies of procedures
are maintained at
local operating stations to assist
operators
in system/component
operation.
The expired procedure
was subsequently
replaced with a
current controlled copy of the procedure.
The inspector toured plant areas
and verified that the procedures
at local operating stations
were properly controlled.
Areas
inspected
included the steam tunnel,
emergency diesel
generator
building, fuel handling building,
and auxiliary control
room.
No
discrepancies
were identified.
Also, the licensee's
periodic
audits of controlled procedures
in these
areas
were reviewed.
The
audit reports indicated that these
procedures
were generally well
maintained.
No violations or deviations
were identified.
Facility Tours
and Observations
Throughout the inspection period, facility tours were conducted to
observe operations,
surveillance,
and maintenance activities in
progress.
Some of these observations
were conducted during
backshifts.
Also, during this inspection period, licensee
meetings
were attended
by the inspectors
to observe
planning
and
management activities.
The facility tours
and observations
encompassed
the following areas:
security perimeter fence;
control
room;
emergency diesel
generator building; reactor
auxiliary building; waste processing
building; turbine building;
fuel handling building; emergency service water building; battery
rooms; electrical
switchgear
rooms; the technical
support center;
and the emergency operations facility.
During these tours, the following observations
were made:
(I)
Monitoring Instrumentation
- Equipment operating status,
area atmospheric
and liquid radiation monitors, electrical
system lineup, reactor operating
parameters,
and auxiliary
equipment operating
parameters
were observed to'verify that
indicated parameters
were in accordance
with the
TS for the
current operational
mode.
Shift Staffing - The inspectors verified that operating
shift staffing was in accordance
with TS requirements
and
that control
room operations
were being conducted
in an
orderly and professional
manner.
In addition, the inspector
observed shift turnover s on various occasions
to verify the
continuity of plant status,
operational
problems,
and other
pertinent plant information during these turnovers.
Plant Housekeeping
Conditions - Storage of material
and
components,
and cleanliness
conditions of various areas
throughout the facility were observed to determine
whether
safety and/or fire hazards
existed.
Radiological Protection
Program - Radiation protection
control activities were observed routinely to verify that
these activities were in conformance with the facility
policies
and procedures,
and in compliance with regulatory
requirements.
The inspectors
also reviewed selected
radiation work permits to verify that controls were
adequate.
Security Control - The performance of various shifts of the
security force was observed
in the conduct of daily
activities which included:
protected
and vital area
access
controls;
searching of personnel,
packages,
and vehicles;
badge
issuance
and retrieval; escorting of visitors;
patrols;
and compensatory
posts.
In addition, the inspector
observed
the operational
status of closed circuit television
monitors, the intrusion detection
system in the central
and
secondary
alarm stations,
protected
area lighting, protected
and vital area barrier integrity,
and the security
organization interface with operations
and maintenance.
Fire Protection
- Fire protection activities, staffing and
equipment
were observed to verify that fire brigade staffing
was appropriate
and that fire alarms,
extinguishing
equipment,
actuating controls, fire fighting equipment,
emergency
equipment,
and fire barriers
were operable.
Between January
and June
1993, significant amounts of
scaffolding had
been erected
in the
190 foot elevation of
the reactor auxiliary building in the vicinity of the "A"
residual
heat removal
(RHR) and containment
spray
pumps,
and
above safety-related
system piping, electrical junction
boxes,
conduits,
and the area
HVAC unit.
The scaffolding
had been erected to support the correction of problems which
have occurred
due to ground water intrusion
(See
NRC
Inspection
Report 50-400/92-04).
'Concurrently
on Hay 19,
1993, the licensee
performed
a maintenance
outage to repair
several
small deficiencies in "B" safety train components
including the "C"- CSIP,
"B" RHR,
CCW,
and
ESW pumps.
The
inspector questioned
the wisdom of placing the "B" RHR
system under
an equipment clearance
which rendered it
inoperable at the
same time that scaffolding was erected
around the "A" RHR pump
and system piping.
The inspector
discussed
this situation with licensee
management.
Although
the licensee
had established
a program for the control of
no seismic evaluations for erected-scaffolding
were required
by the implementing procedure.
The inspector reviewed
PLP-401
(Ladder, Scaffold,
and
Equipment
Use
and Storage)
and noted specific exclusions for
the erection of scaffolding over redundant safety trains
simultaneously.
The intent of this step
was to prohibit the,
loss of both safety trains due solely to the failure of
Regulatoi y Guide 1.29 discusses
the "two-over-
one" situation where nonsafety-related
structures
whose
continued function is not required,
but whose failure could
reduce the functioning of any safety-related
equipment,
should
be seismically designed
and constructed.
The
inspector
noted that procedure
PLP-401 did not require
a
seismic evaluation of erected scaffolding over safety-
related
components;
particularly when scaffolding is erected
over the only available train of safety-related
systems,
as
was the case
on May 19.
The inspector related this concern to licensee
management,
who concurred with the inspectors
comments
and directed
a
seismic evaluation
be performed for the "A" RHR and
- containment
spray
pump room.
This evaluation
concluded that
the scaffolding would not damage safety-related
components
during
a seismic event.
While the evaluation yielded favorable results for the
scaffolding erected
over the "A" RHR system,
the inspectors
concluded that the licensee's
controls over this situation
were weak, in that the evaluation
should
have
been
done
prior to the removal of the "B" RHR system from service
on
Nay 19.
The inspectors
noted that the licensee
has taken
administrative
measures
to ensure that scaffolding is
no'onger
erected
over safety-related
equipment until the
procedure is enhanced
to include better controls.
The inspectors
found plant housekeeping
and material condition of
components
to be satisfactory.
The licensee's
adherence
to
radiological controls, security controls, fire protection
requirements,
and
TS requirements
in these
areas
was satisfactory.
Review of Nonconformance
Reports
Adverse Condition Reports
(ACR) were reviewed to verify the
following:
TS were complied with, corrective actions
and generic
items were identified and items were reported
as required
by
(2)
ACR 93-220 reported that the electrical
containment
penetration for the Integrated
Reactor Vessel
Head
(IRVH)
bridge hoist was not provided with a secondary
protection device
as specified in plant control wiring
drawings. 'uring
a walkdown of these circuit breakers
used
to provide overcurrent protection
on Hay 25,
1993, the
licensee
discovered that only one breaker provided power to
the bridge hoist located inside containment.
The other
breaker
specified
on the dr awings
was found to be
disconnected
from the circuit.
When notified of this
condition, plant operators
deenergized
the circuit by
establishing
an equipment clearance
in accordance
with TS 3.8.4.l.a.
The inspector verified the clearance
tagout in
the field.
Although the tags were found to be on the
specified
components,
the inspector noted that
one of the
breakers. which were tagged
(1D12-7BR) was labeled
as
a
"spare" instead of a power source for the bridge hoist.
This spare
breaker coincided with the one which was not
terminated to the circuit.
The licensee's
investigation
into this matter concluded that this wiring deficiency
had
existed since initial plant construction.
The licensee
has experienced
previous
problems with these
types of electrical penetrations.
On Harch 14,
1990,
licensee
personnel
discovered
a discrepancy
in the amperage
rating for a circuit breaker providing overcurrent
protection.
The licensee
issued
LER 90-08 reporting this
condition.
The corrective actions associated
with this
LER
included
a review of the calculations of similar circuits
. and
a field verification that breakers
used
as protection
, devices
were properly identified/verified.
These actions
were completed in Nay 1991.
This matter was considered
to
be
a licensee identified violation in NRC Inspection
Report
50-400/90-06
(NCV 400/90-06-01).
Although the condition reported
by ACR 93-220 was identified
by the licensee,
corrective actions
performed for the
previous deficiency should
have identified the labeling
problem for breaker
1D12-7BR and the inability of this
secondary
breaker to perform its function.
Therefore, this
matter is considered
to be
a violation of TS 3.8.4.a.
Violation (400/93-12-01):
Failure to maintain the
IRVH
bridge hoist electrical
containment penetration
conductor
overcurrent protective device operable.
ACR 93-205 reported that on Hay 17,
1993, the
RAB normal
exhaust
fan E-18 could not be secured
from the control
switch in the main control room.
Auxiliary operators
noted
locally that the fan's circuit breaker
was still closed
and
an electrical
burning smell
was present.
The breaker
subsequently
automatically tripped open after two
unsuccessful
manual
attempts.
Troubleshooting of this Asea
Brown Boveri LK-16 circuit breaker revealed that the trip
coil had burned
up and the breaker
had mechanically
bound in
the, closed position.
Although this application of the LK-16
breaker
was in nonsafety-related
equipment,
many safety-
related
components
receive electrical
power through
LK-16
breakers.
Several
previous failures of these
types of breakers
have
occurred in the past.
As reported in LER 92-09,
a failure
of the E-18 supply breaker contributed to a reactor trip on
July 15,
1992.
As mentioned in NRC Inspection
Reports
50-
400/92-26,
50-400/92-13,
and 50-400/90-13,
the licensee
has
performed
numerous corrective actions to improve breaker
reliability including breaker
physical modifications
(PCR-
3510,
480 volt Drawout Breakers
used
as Contacters)
and
periodic preventive maintenance
checks.
Although this
action reduced
the number of breaker failures, additional
failures
have occurred
even
on the modified breakers
(the E-
18 breaker
had
been previously modified in November 1990).
The licensee
plans to replace this switchgear beginning in
the next refueling outage
scheduled for March 1994
(PCR-
6526,
Frequently Cycled
LK Breakers;
PCR-6714,
Frequently
Cycled Non-Safety
LK Breakers;
PCR-6715, Train "B" Load
Centers
LK Breakers
Replacement;
and
PCR-6896,
Non-Safety
Train "B" and General
Services
Bus Section
1 Breaker
Replacement).
Until then the licensee
developed
interim
corrective action.
The licensee
determined that
a
significant number of the failures were from a few heavily
cycled breakers.
These
breakers
are being identified and
"retired".
In addition, the preventive maintenance
frequency for the heavily cycled breakers will be increased
to ensure
breaker reliability.
On May 30,
1993, while attempting to deenergize
nonsafety-
related
MCC 1-481 to perform LK-16 circuit breaker
modifications per PCR-3510, plant operators
could not open
the 480 volt feeder breaker for this bus.
Alternatively,
the operators
opened
the corresponding
6.9
KV supply breaker
to the
MCC.
Troubleshooting of this Asea
Brown Boveri LK-42
feeder breaker revealed similar problems to those
found on
the LK-16 circuit breakers.
The breakers
are very similar
in design.
The licensee
had already increased
the scope of
modification PCR-3510 to include all the
LK breakers (i.e.,
LK-8, LK-16, LK-25, LK-32 and LK-42).
Approximately 200 of
the
243
LK circuit breakers
have
been modified.
The
modification was subsequently
performed
on this feeder
breaker.
The failure of this circuit breaker differ ed from
previous failures in that the feeder breaker did not
experience
many cycles
and was not anticipated to exhibit
the
same
problems
as the heavily cycled breakers.
As
discussed
in paragraph
6, another failure mechanism for LK
breakers
involves the closing coils.
The inspectors will
continue to monitor the licensee's
progress
in correcting
these
breaker problems.
ACR 93-237 reported that the lock for the circuit breaker
for valve 1CT-95, Containment
Spray
Pump
1B-SB Recirculation-
Valve, was not installed
as required
by Operations
Management
Manual procedure
OMM-011, Control of Locked
Valves.
Valve 1CT-95 is
a
normally closed motor-operated
containment isolation valve whose breaker is required to be
locked/off during normal plant operating conditions,
as
described
in the plant
FSAR. 'uring performance of
surveillance
procedure
OST-1119,
Containment
Spray
Operability Train
B quarterly Interval,
on June
2,
1993,
operators
were required to remove the lock and close the
breaker.
This would allow for the manipulation of valve
1CT-95 to support surveillance testing of the "B"
containment
spray
pump.
Section 7.3, step
48 of the
procedure
requires
operators
to turn off and lock the
breaker to valve
1CT-95 following the
pump test.
There are
two sign-off steps
in the procedure for this action,
one for
the operator performing the step
and another for an
independent verifier.
Interviews with'he involved
personnel
revealed that while performing the step,
the
operator turned the breaker off but failed to re-install the
lock as required.
Additionally, both sign-off steps
were
initialed by the operator
and the independent verifier
indicating that the lock had
been installed
and verified to
be in place.
The individuals involved indicated that the
error was due to oversight
and that they simply did not read
the procedure
step completely.
The inspectors
concluded
that this was not
a case of record falsification.
While this incident was identified by the licensee, it is
the latest of several
examples of plant performance
where
either self-checking or independent verification were
lacking.
On April 28,
1993, operations
personnel
went to
the wrong breaker cubicle during efforts to swap safety
trains.
This resulted in an inadvertent start of the "B"
CCW pump.
In the past,
NRC violations have
been
issued
(Violations 400/91-27-01
and 400/91-09-01) for incidents
involving inadequate
self-checking or independent
verification.
One involved a mispositioned
NI comparator
switch which rendered
the
gPTR alarm inoperable in January
1992.
The other involved instrument process
tubing for,a
differential pressure
switch in the
EDG fuel oil system,
which was neither
reconnected
nor properly verified to be
connected prior to returning the instrument to service in
April 1991.
This latest
example involving the failure to
properly implement procedure
OST-1119 is contrary to the
requirements 'of TS 6.8.l.a
and is being cited because
violations have
been previously identified by both the
NRC
and the licensee
in the areas of procedural
adherence
and
independent verification, indicating
a need for additional
management
attention in those
areas.
Violation (400/93-12-02):
Failure to properly implement
plant procedures
as required
by TS 6.8. I.a.
3.
Surveillance Observation
(61726)
Surveillance tests
were observed to verify that approved
procedures
were
being used; qualified personnel
were conducting the tests;
tests
were
adequate
to verify equipment operability; calibrated
equipment
was
utilized;
and
TS requirements
were followed.
The following tests
were
observed
and/or data reviewed:
OST-1007
OST-1039
OST-1079
OST-1081
HST-I0137
HST- I0138
HST- I0151
EST-222
EPT-126T
CVCS/SI System Operability quarterly Interval
Calculation of quadrant
Power Tilt Ratio, Meekly
Interval
(With Alarm Operable)
Containment Isolation Valves ISI Test quarterly
Interval
Containment Visual Inspection
When Containment
Integrity is Required
Hain Steam/Feedwater
Flow Loop 2 (F-0485/F-0486)
Operational
Test
Hain Steam/Feedwater
Flow Loop 3 (F-0494/F-0497)
Operational
Test
C Narrow Range
Level
Loop (L-0496)
Operational
Test
Procedure for the Type
B LLRT of the Personnel
Air
Lock Barrel
Turbine Volumetric Test
4.
The performance of these
procedures
was found to be satisfactory with
proper
use of calibrated test equipment,
necessary
communications,
established,
notification/authorization of control
room personnel,
and
knowledgeable
personnel
having performed the tasks.
The inspectors
noted good use of the plant simulator for performing practice runs
on
procedure
EPT-126T.
", This allowed operators
to smoothly execute
the
actual test,
which was
an infrequently used procedure that had the
potential for introducing unwanted plant transients if not performed
correctly.
No violations or deviations
were observed.
Haintenance
Observation
(62703)
The inspector observed/reviewed
maintenance activities to verify that
correct equipment clearances
were in effect; work requests
and fire
prevention work permits were issued
and
TS requirements
were being
followed.
Haintenance
was observed
and work packages
were reviewed for
the following maintenance activities:
~
Preventive
maintenance
on the startup transformer feeder breaker
to auxiliary bus
1E (breaker
121) in accordance
with procedure
PH-
E006,
6.9
KV 3000
Amp Air Circuit Breaker
PM.
~
Repair nitrogen leak on the actuator for the "A" main feedwater
isolation valve
1FW-159 in accordance
with procedure
CM-H0059,
Isolation Valve Actuator Disassembly
and Maintenance,
Pilot Check Valve Shim Replacement
and Fill/Bleed Procedure.
~
Replacement
of bent yoke that actuates circuit breaker
122
mechanism
operated cell switch.
~
Troubleshooting/calibration
of the nitrogen pressure
switch for
the "C" main feedwater isolation valve
1FW-217 in accordance
with
procedure
PIC-I100,
Pressure
and Differential Pressure
Switch
Inspection
and Calibration.
~
Repair of the "A" steam generator
blowdown flow control valve
1BD-
8 in accordance
with procedure
CM-H0033,
ITT 'Hamel
Dahl
V500
Series
Cage Trim Valves Disassembly
and Maintenance.
4
~
Troubleshooting
the failure of the "B" chilled water
pump breaker
to close in accordance
with procedures
EPT-033,
Emergency
Safeguards
Sequencer
System Test,
and
PH-E0012,
480
VAC Load
Center Breaker
and Cubicle
PH.
The performance of work was satisfactory with proper documentation of
removed
components
and independent verification of the reinstallation.
No violations or deviations
were identified.
Safety Systems
Walkdown (71710)
The inspector
conducted
a walkdown of portions of the fire protection
system to verify that the lineup was in accordance
with license
requirements
for system operability and that, the system drawing and
procedure correctly reflected "as-built" plant conditions.
The inspector noted that the material condition of components
in the
valve pit area adjacent to the motor driven fire, pump was poor.
The
corresponding
area near the diesel
engine driven fire pump was in a much
better material condition.
The inspector discussed this observation
with licensee
personnel
and discovered that efforts were already
underway to upgrade
the material condition of the motor driven fire pump
components
as
had been recently done to the engine driven fire pump
components.
r
From a review of plant drawings the inspector
noted that no drawing
existed for the fuel oil supply tank and delivery system to the diesel
engine driven fire pump.
Although only a few valves existed in this
system
(one flow isolation valve,
one tank drain valve,
and two level
indicator isolation valves),
the inspector considered
the lack of a flow
10
drawing to be unusual.
Licensee
personnel
stated that they would review
this matter to determine if a drawing was appropriate.
No violations or deviations
were identified.
Followup of Onsite Events
(93702)
At 5:32 p.m.,
on Hay 23,
1993, the
1B-SB emergency
bus
was inadvertently
deenergized
which caused
an automatic start of the "B" emergency diesel
generator.
A power increase
was in progress
follow'ing the outage.
When
operators
attempted to transfer
AC electrical
power from the Startup
Transformers
(SUT) to the Unit Auxiliary Transformers
(UAT) a circuit
breaker malfunctioned which caused auxiliary bus lE to be supplied from
both the
UAT and
SUT.
Operators
proceeded
to manually open the
feeder breaker
(121) to restore
the electrical lineup to the desired
configuration.
The auxiliary contacts
associated
with the
UAT feeder
breaker
(122) were not properly engaged
due to a damaged
actuating
mechanism
and provided
a false breaker
open signal to the
emergency
bus feeder breaker
(125)
even though breaker
122 was actually
closed.
When breaker
121 was manually opened,
feeder breaker
125 was
provided with open signals
from both breakers
121'nd
122 and
automatically
opened
as designed.
This action deenergized
the
emergency
bus
and the emergency diesel
generator
started
on
a low bus
voltage signal.
Haintenance
was performed
on breaker
122 to replace the
actuating
mechanism for the auxiliary contacts
which were found to be
bent.
The licensee
believes
the damage to the actuating
mechanism
occurred
when the breaker
was previously racked in during the October
1992 refueling outage.
At 7:54 a.m.,
on Hay 24, the emergency diesel
generator
was secured
and the electrical distribution lineup returned to
the desired configuration with the
UAT supplying auxiliary bus lE.
During this event,
system actuated
as designed
and
a containment ventilation isolation signal
was generated
due to
losing power to the containment
system radiation area monitors.
With
the exception of the "B" chilled water
pump which failed to start, plant
equipment
operated
properly.
The licensee
investigated
the problem and
determined that the closing coil on the LK-16 circuit breaker for this
pump had failed.
The closing coil was subsequently
replaced
and the
pump was tested satisfactory.
The licensee
has experienced
several
previous
problems with proper
engagement
of 6.9
KV breaker
secondary
contacts.
As discussed
in NRC
Inspection
Report 50-400/92-04,
closing problems
have
been identified
for these
types of breakers.
On October 1,
1988,
a similar event to
this one occurred.
While in a refueling outage,
power was inadvertently
interrupted to the lA-SA emergency
bus during an attempt to realign-the
electrical distribution system to backfeed electrical
power through the
In this event the
SUT feeder breaker to auxiliary
bus
1D failed to open
when the
UAT feeder breaker
was closed.
'When
operators
manually opened the
SUT feeder breaker the
1A-SA feeder
breaker automatically
opened
and deenergized
the emergency
bus.
Troubleshooting for this condition also revealed that the actuating
mechanism for the auxiliary contacts
on the
UAT feeder breaker
were
bent.
Corrective action to prevent recurrence of this event included
a
repair to the actuating
mechanism,
and procedural
guidance
provided to
operators
to prevent actuating
mechanism
damage
when racking 6.9
KV
breakers
into the connected position.
Operating procedure
Electrical Distribution,
wa's revised to provide
a note alerting
operators
to visually ensure that the cell switch is properly aligned
with the actuating
mechanism.
The failure to properly implement procedure
OP-156.02 (i.e., verify the
alignment of the circuit breaker actuating
mechanism with the cell
switch when the breaker
was placed into service during the previous
refueling outage)
is contrary to the requirements
of TS 6.8. l.a and is
considered to be another
example of the violation discussed
in paragraph
2.c.(3) of this report.
Outage Activities (71707)
Hajor work performed during this planned
outage
included repairs to
blowdown valve IBD-8, oil addition to the "B" reactor
coolant
pump, repair, of nitrogen leaks
on the operators for valves
1FW-
159 and
1FW-217, replacement of the motor for the "A" heater drain pump,
and repair of through-wall leaks
on the seal
water supply piping for the
"B" condensate
booster
pump.
As mentioned in
NRC Inspection
Report 50-400/93-08,
containment
inleakage
had increased
and
a temporary modification had
been installed
to valve
the valve and reduce
secondary
system
inleakage to the
sump.
During this inspection period,
a gradual
increase
in sump inleakage
occurred.
Licensee
management
decided to
take the unit off line and repair this valve prior to the upcoming
summer peak load period.
Significant planning activities included the use of a spare
"mockup"
valve
and actuator for the lBD-8 work and
a spare reactor, coolant
pump
motor for the oil addition.
The simulator was uti'iized to provide
operators
with unit shutdown
and startup training,
as well as potential
which could be expected
during these plant evolutions (i.e.,
and premature reactor criticality).
The inspector
found this pre-planning to be effective and the actual evolutions were
performed without mishap.
Additional activities performed
by the licensee
during the outage
included
a containment visual inspection to identify fibrous air filters
or other sources of fibrous material in accordance
with NRC Bulletin 93-
02, Debris Plugging of Emergency
Core Cooling Suction Strainers,
and
a
visual inspection of the reactor coolant pressure
boundary for boric
acid leakage.
No fibrous material
and only minor boron residue
was
found.
Work tickets were generated
to correct the boric acid leakage
during the next refueling outage.
12
The inspector
requested
the documented
inspection results for the boric
acid leakage
inspection.
The inspector
was informed that the results
were not well documented.
Procedure
PLP-600,
Boron Corrosion
Program,
and inspection
procedure
OPT-1519,
Containment Visual Inspection for
Boron Leakage
Every Outage
Shutdown, specified that
a walkdown
inspection
be performed prior to cooldown in Hode 3.
Since
a plant
cooldown was not commenced,
the inspection procedure,
which included
an
attachment
specifying areas
to be inspected
and results,
was not
implemented.
After the inspector discussed this matter with licensee
management,
the data sheets
were filled out with the results for the
areas
inspected.
The inspector considered
the licensee's
documentation
of the boric acid leak inspection to be poor.
Adverse Weather Operations
(71707)
On June 4,
1993,
a tornado warning was issued for Wake County.
The
inspector reviewed the licensee's
preparations
performed in response
to
this warning.
No tornadoes
approached
the plant
and
no plant damage
occurred during the adverse
weather conditions.
The inspector reviewed
the provisions contained in the licensee's
emergency
plan for handling
these situations
and also reviewed procedure
AP-301, Adverse Weather
Operations.
The licensee's
emergency
plan requires
the declaration of
an Unusual
Event if a tornado crosses
the exclusion area
boundary,
and
if sustained
wind speeds
exceed
90
HPH then
an
ALERT would be declared.
If a tornado
impacts the power block and sustained
winds exceed
100
HPH
then
a site emergency
would be declared.
Procedure
AP-301 contains
provisions to ensure
the plant is placed in hot standby
(Hode 3) at
least
two hours prior to the anticipated arrival of sustained
winds in
excess
of 73 HPH.
For the tornado warning issued
on June 4, loose
material
was
removed from exposed
areas,
acc'ess
doors closed,
safety
lines fixed,
and emergency
equipment
was verified to be available.
The
inspector
found that the licensee's
emergency
plan and procedures
were
properly implemented.
Review of Licensee
Event
Repo} ts (92700)
The following LERs were reviewed for potential generic
impact, to detect
trends,
and to determine whether corrective actions
appeared
appropriate.
Events that were'reported
immediately were reviewed
as
they occurred to determine if the
TS were satisfied.
LERs were reviewed
in accordance
with the current
a ~
(Open)
LER 93-05:
This
LER reported
an entry into TS 3.0.3
on
April 28,
1993 during an in-service stroke test
on two CSIP
discharge
cross
connect valves.
As discussed
in NRC Inspection
.
Reports
50-400/93-08
and 50-400/93-10,
the "B" CSIP rotating
assembly
was replaced following a shaft failure.
Following the
return to service for this pump, valves'CS-218
and
1CS-220 were
stroke tested to satisfy
IST requirements.
These valves isolate
the "A" CSIP from the normal high head safety injection flow path
when closed.
The valves'troke
times were approximately ten
seconds
each
and they were closed for a total of less than
one
13
b.
minute.
On April 29, it was determined that the "B" CSIP had not
yet been demonstrated fully operable
and that both
ECCS flow paths
were inoperable for the seconds
that the valves were closed.
This
constituted
a TS 3.0.3 entry.
The "C" CSIP was subsequently
placed into service
and the
"B", CSIP was declared
The licensee attributed this event to several
causes,
all stemming
from personnel
error.
Inadequate
evaluation
and review of the "B"
CSIP test data coupled with a lack of understanding
on the use
and
restrictions of the
JCO process
were cited as reasons
for the
premature declaration of the
pump to be operable.
The licensee
will train appropriate plant personnel
on this event
and the
proper
use of JCOs,
Engineering Evaluations,
Technical
Specifications,
and other processes
applicable to this event.
Additionally, the "B" CSIP will now remain inoperable until
conditions permit full flow testing to verify the pump's ability
to meet vendor
and
TS runout limits.
This
LER will remain
open
pending completion of the above actions.
(Open)
LER 93-06:
This
LER reported
two separate
incidents
where
the requirements
of TS 4. 11. l. 1. 1 were not met.
Specifically,
on
Hay 17 and Hay 31,
1993, it was determined that the automatic
sampling device
on the Secondary
Waste
Sample
Tank
(SWST) effluent
line had malfunctioned during releases
from the
SWST.
Technical
Specification'.11. 1.1.1 requires
continuous
sampling of the
SWST
effluent during releases
to provide
a weekly radioactivity
composite.
The composite
sample
would be used
as
a backup to the
normal radiation monitor in the event it became
This
composite
sample is typically discarded
when the radiation monitor
is operable.
In both instances
the
SWST releases
were immediately
secured
and the in-line radiation monitor was used to verify that
no detectable
levels of radioactivity existed.
The sampling device malfunctions were caused
by a failure in, the
unit's electronic counter .
After the second failure, the licensee
consulted with the vendor who attributed the cause of both
failures to inadequate
surge protection in the counter
power
supply.
Immediate corrective actions
were taken to replace the
automatic
sampling device with a modified version which included
enhanced
surge protection for the counter power supply.
Long-term
corrective actions
are to include developing test procedures
to
more thoroughly test
or calibrate the composite
sampling units.
This
LER will remain
open pending completion of the procedure,
development.
10.
Licensee Action on Previously Identified Inspection
Findings
(92702
K
92701)
(Closed)
URI 400/93-10-01:
Review calculations
and licensee
actions
regarding the seismic qualification of AFW piping.
0
14
As discussed
in NRC Inspection Report 50-400/93-10,
the inspectors
identified two valves in the
AFW system,
1AF-5 and lAF-24, which
appeared
to be installed without adequate
supports.
These two-inch
motor-operated
valves are in the recirculation lines from the two motor-
driven
AFW pumps to the condensate
storage
tank..
The valve bodies
are
installed horizontally in the piping with big Limitorque motor operators
attached
to their sides.
The motor operators
had
no additional
supports
other than the small two-inch piping,
and
seemed
to impose
a significant
bending
moment
on that piping.
The inspectors
brought this observation
to the licensee's
attention
who researched
the original design
calculations for that section of piping and discovered
a modeling error
in the centers of gravity for the two valves.
Specifically, Stress
Analysis Calculation 71-1,
"AF Piping From Mass Points
66 and 453 at
Floor (flevation 261') to Steam Generator Auxiliary Feedwater
Pumps
Discharge
Nozzles
and Anchor Point 4903 at Wall", incorrectly applied
the mass of the composite valve body/motor operator
assemblies
at the
center of gravity of the valve bodies
themselves
for valves
lAF-5 and
IAF-24, and not at the cente} of gravity of the composite valve
assemblies.
Stress
calculations resulting from this modeling error
incorrectly yielded stress
figures which wer e within the design
basis
allowable stress criteria as referenced
in the
FSAR, Table 3.9.3-11.
Recalculations
of the actual
pipe stresses
imposed
by the unsupported
motor operators yielded stress
values of approximately
33 ksi.
This
figure exceeded
the allowable value for the emergency condition (which,
as defined
by FSAR Table 3.9.3-7,
included the safe
shutdown earthquake
in its loading combination)
by 6 ksi.
This rendered that section of the
seismic class I AFW system outside of its original design allowable
str ess limits.
Upon discovering this, the licensee
performed
an
immediate operability determination
as allowed by the guidelines of NRC Generic Letter 91-18, "Information to Licensees
Regarding
Two NRC
Inspection
Manual Sections
on Resolution of Degraded
and Nonconforming
Conditions
and
on Operability" and the licensee's
own Design Guide No.
DG-II.20,
Design Guide for Civil/Structural Operability Reviews".
The
design guide incorporated
more recent rules for calculating .stress
limits as specified in Appendix
F of the Appendices to ASME Boiler &
Pressure
Vessel
Code,Section III, 1986 edition.
Specifically,
a stress
limit of two times the material yield stress. of 36 ksi, or 72 ksi, could
be used in determining the short-term structural integrity of the
recirculation piping.
Based
on the actual
stresses
being calculated at
33 ksi, the system
was considered
operable in the short-term.
Stress
Analysis Calculation 71-1, which contained
the center of gravity
modeling error was originally developed in 1974 by the plant,'s architect
engineer.
The calculation received
reviews by the licensee
in 1986 in
accordance
with the guidelines established
"Seismic Analyses for As-Built Safety-Related
Piping Systems"..
However,
those reviews failed to identify the modeling'rror which resulted in
that section of AFW piping being outside of its design basis.
The
failure to perform adequate
measures
to ensure
the accuracy of design
with respect to this seismic class I system is considered to be
a
violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control.
15
Violation (400/93-12-03):
Failure to establish
adequate
measures
to
verify that designs
were technically accurate
with respect to the design
basis for the
AFW system.
In accordance
with the requirements
of NRC Generic Letter 91-18 and the
licensee's
design guide,
an engineering evaluation which will document
the current,
short-term operable condition,
as well as
a modification
which will restore the system to long-term acceptable
status,
are being
developed
by the licensee.
Exit Interview (30703)
The inspectors
met with licensee
representatives
(denoted in paragraph
1) at the conclusion of the inspection
on June
21,
1993.
During this
meeting,
the inspectors
summarized
the scope
and findings of the
inspection
as they are detailed in this report, with particular emphasis
on the Violations addressed
below.
The licensee representatives
acknowledged
the inspector's
comments
and did not identify as
proprietary
any of the materials
provided to or reviewed
by the
inspectors
during this inspection.
No dissenting
comments
from the
licensee
were received.
Item Number
Descri tion and Reference
400/93.-12-01
400/93-12-02
400/93-12-03
and Initialisms
VIO:
Failure to maintain the
IRVH bridge
hoist electrical
containment. penetration
conductor overcurrent protective device
paragraph 2.c.(l).
VIO:
Failure to properly implement plant
procedures
as required
by TS 6.8.l.a.,
paragraphs
2.c.(3)
and 6.
VIO:
Failure to establish
adequate,
measures
to verify that designs
were
technically accurate with respect to the
design basis for the
AFW system,
paragraph
10.
ACR
ASNE
CFR
CSIP
EPT
Adverse Condition Report
Administrative Procedure
American Society of Hechanical
Engineers
Component Cooling Water
Code of Federal
Regulations
Charging Safety Injection
Pump
Chemical
Volume Control System
Emergency
Core Cooling System
Engineering
Performance
Test
EST
IRVH
JCO
KV
KSI
LER
HCC
HPK
NRC
OPT
OST
PLP
PH
SWST
TS
VAC
16
Engineering Surveillance Test
Emergency Service Water
Final Safety Analysis Report
Heating, Ventilation and Air Conditioning
Integrated
Reactor
Vessel
Head
Inservice Inspection
Inservice Testing
Justification for Continued Operation
Kilovolt
Kilopounds per square
inch
Licensee
Event Report
Local Leak Rate Test
Hotor Control Center
Miles Per Hour
Non-Cited Violation
Nuclear Regulatory
Commission
Operations
Performance
Test
Operations
Surveillance
Test
Plant
Change
Request
Plant Procedure
Preventive
Maintenance
Quadrant
Power Tilt Ratio
Reactor Auxiliary Building
Residual
Heat
Removal
Safety Injection
Startup Transformers
Secondary
Waste
Sample
Tank
Technical Specification
Unit Auxiliary Feedwater
Unresolved
Item
Volt Alternating Current
Violation