ML18012A426
| ML18012A426 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/07/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18012A425 | List: |
| References | |
| 50-400-96-09, 50-400-96-9, NUDOCS 9611180109 | |
| Download: ML18012A426 (30) | |
See also: IR 05000400/1996009
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION II
Docket No:
License
No:
50-400
Report
No:
50-400/96-09
Licensee:
Carolina Power
8 Light (CPKL)
Facility:
Shearon Harris Nuclear Power Plant, Unit 1
Location:
5413 Shearon Harris Road
New Hill, NC 27562
Dates:
September
1
- October
12,
1996
Inspectors:
J.
Brady, Senior Resident
Inspector
D. Roberts,
Resident
Inspector
F. Jape,
Senior Project Hanager
(H6.1)
Approved by:
H. Shymlock, Chief, Projects
Branch 4
Division of Reactor
Projects
Enclosure
2
961ii80i09 9bii07
ADQCK 05000400
8
EXECUTIVE SUMMARY
Shearon Harris Nuclear
Power Plant, Unit 1
NRC Inspection Report 50-400/96-09
This integrated inspection included aspects of'icensee
operations,
engineering,
maintenance,
and plant support.
The report cover s
a six-week
period of resident inspection;
in addition, it includes the results of an
announced
inspection
by a regional project manager.
~0erations
Operator performance during
a plant trip and reactor startup
was good
(Sections
01.2 and 01.4).
-A violation with two examples of failure to
follow procedures
was identified.
One involved the inadvertent dilution
of the
RWST (Section 04.1),
and the other involved the use of a non-
approved procedure resulting in the inadvertent actuation of an
emergency ventilation fan in the fuel handling building (Section 04.2).
Self assessment
activities by the Nuclear Assessment
Section were good,
including one that resulted in the identification of a missed Technical
Specification surveillance
requirement
(Sections
07.1 and M8.7).
The licensee
reasonably
addressed
corrective actions for problems
identified during the period;
however those identified in Licensee
Event
Report 96-014 did not address
a 1989 inadequate reportability
determination
(Section 08.2).
Haintenance
~
One violation was identified for failure to properly implement the
maintenance
rule,
10 CFR 50.65 (Section H8.8).
The performance of on-
line maintenance
was found to be within the expectations
of NRC guidance
(Section M6.1) .
~
Several
Licensee
Event Reports were generated
involving Technical
Specification surveillance
implementation problems
(Section M8.1).
En ineerin
An unresolved
item was identified for the use of administrative
means
(Technical Specification Interpretation procedures)
in lieu of a license
amendment in relation to the ultimate heat sink (Section E1.2).
This
unresolved
item also involved discrepancies
between the Final Safety
Analysis Report
and parameters
used in defining the plant's design basis
as it relates to the ultimate heat sink.
A previous condition determined to be outside the plant's design basis
was identified in relation to non-seismic piping systems
being connected
to the Refueling Water
Storage
Tank (Section E8.1).
0
Licensee
personnel
appropriately
responded to Hurricane Fran and made
the proper
NRC notifications when emergency
response
capability was
hampered
and equipment
became unavailable
(Section
P2. 1).
One violation was identified f'r failure to close
a protected
area
boundary gate following vehicle access
(Section S1.2).
Re rt Details
Summar
of Plant Status
Unit 1 began this period operating in Hode
1 at 100 percent
power.
On
September
3,
1996, operators
manually tripped the reactor after receiving
indications of a loss of normal service water
(NSW).
The operating
"B" NSW
pump experienced
a sheared
shaft
and repeated
attempts to start the standby
"A" pump were unsuccessful.
The loss of this non-safety related
system
required that the plant be manually tripped per abnormal
operating procedures
in order to protect
numerous
secondary plant components that receive cooling
water from NSW.
Attempts to determine
why the "A" pump tripped on the initial
start attempt were unsuccessful;
however,
a temporary modification was
installed to monitor the pump's logic circuitry during future operation.
The
second
and third attempts apparently failed due to a timed protective
interlock of which operators
were unaware.
The plant remained
shutdown in Hode 3, Hot Standby,
through September
5 when
Hurricane Fran passed
near the site.
Electrical outages
caused
by the
hurricane resulted in the unavailability of approximately
90 percent of the
evacuation
sirens located in the Emergency Planning Zone.
As a result,
restart of the unit required the joint approval of the Federal
Emergency
Hanagement
Agency and the
NRC.
This approval
was granted
on September
9 and
the unit was restarted
at 10:32 p.m. that night.
The generator
was
synchronized to the grid at 3:42 a.m.
on September
10,
1996,
and the reactor
reached
100 percent
power on September
11,
1996.
On October 6,
1996, reactor
power was reduced to approximately 53 percent
following the loss of main feedwater
pump "B".
The pump was secured after
operators
noticed
a leaking outboard seal
and
a considerable
amount of oil and
water in the oil sump.
After repairs to the pump, reactor
power was returned
to 100 percent
on October 10, 1996.
01
Conduct of Operations
01.1
General
Comments
71707
I. 0
rations
Using Inspection Procedure 71707, the inspector s conducted
frequent
reviews of ongoing plant operations.
In general,
the conduct of
operations
was professional
and safety-conscious.
Specific events
and
noteworthy observations
are detailed in the sections
below.
01.2
Onsite
Res onse to Events
a.
Ins ection Sco
e
93702
The inspector
responded to the plant upon notification of a manual reactor trip on the night of September
3,
1996.
The plant was already
stabilized in Hode 3 when the inspector
arrived in the control
room.
The inspector
assessed
operator
performance
and plant response.
The
inspector
remained at the site to verify that the plant was stable
and
that the licensee
was addressing
the cause of the trip.
Observations
an
Findin s
On September
3,
1996 at approximately 11:42 p.m., operators initiated a
manual reactor trip after receiving indications of a total loss of
normal service water
(NSW).
The plant was tripped in accordance
with
abnormal
operating procedures
in order to protect secondary
system
components that receive cooling water from this non-safety related
service water system.
All control rods inserted into the core
and other
systems
responded
as expected,
including the standby train of the
emergency
(ESW) system which started
ressure.
system automatically started
on low-
ow steam generator level,
due to the expected shrink in level following
the trip.
Operators
made the appropriate 4-hour notification to the
NRC
in accordance
with 10 CFR 50.72.
A 30-day Licensee
Event Report
(LER)
was submitted in accordance
with 10 CFR 50.73
and is briefly discussed
in section 08.3 of this report.
Licensee
personnel
determined the root cause of the trip to be
a
mechanical
failure of the "B" NSW pump combined with the failure of the
"A" NSW pump to remain running once manually started.
The "B" NSW
pump's impeller shaft sheared.
The discharge
valve for the "B" NSW pump
would not seal
and resulted in having to drain the service water header
to facilitate maintenance.
The valve was replaced,
the header filled,
and the "A" NSW pump started.
The "B" pump was disassembled
and the
shaft was sent to the corporate metallurgy lab for initial analysis.
A
preliminary report determined that the failure was not likely due to the
original manufacturing process.
At the end of the inspection period,
a
final metallurgical analysis
was in progress to determine the failure
mode.
The cause of the "A" pump's initial tr ip was not determined.
Two
subsequent
attempts to start the "A" pump in immediate succession
had
also failed.
The licensee's
root cause investigation determined that
a
timed protective interlock in the pump's starting logic prevented the
second
and third start attempts.
Control
room operators
were unaware of
this pt otective interlock at the time of the event.
This was determined
to be
a training issue.
The
NSW system operating procedures
were
revised to include this information for future reference.
The licensee
planned to incorporate this item along with the results of a pending
investigation into the initial pump trip into the operator training
progr am.
Troubleshooting
determined that
a common-mode failure did not exist for
the 'other
pump.
The "A" pump was returned to service
on September
7,
1996.
The licensee
completed
a safety analysis
using probabilistic risk
assessment
techniques to determine
compensatory
measur es for returning
the plant to service with only one
NSW pump available.
The reactor
was
restarted
on September
9,
1996 and the main generator
was synchronized
to the grid on September
10,
1996.
c.
Conclusions
The inspector
concluded that operator
performance
was satisfactory
during this event.
The licensee's initial efforts to determine
causes
for the
NSW pump problems were. commendable.
01.3
Post- Tri
Review
a.
Ins ection Sco
e
93702
The inspector
reviewed the licensee's
Post-Trip Review report for the
September
3 manual reactor trip to determine whether the licensee
thoroughly assessed
operator
and plant response
and made
sound
recommendations
for corrective actions where needed.
b.
Observations
and Findin s
In accordance
with procedure
OHH-004, Post-trip/Safeguards
Actuation
Review, Revision 7, plant personnel
completed the formal data collection
and investigation process for the September
3rd reactor trip.
The
inspector
reviewed instrument recorder graphs,
plant computer data,
historical
alarm printouts,
and personnel
statements
related to the
trip.
The reviewed data substantiated
the licensee's
assessment
of the
direct cause of the manual trip (loss of normal service water).
The
inspector independently
concluded that plant systems
responded
as
required.
Reactor coolant system temperature
cooled to 545 degrees
Fahrenheit
(F),
an improvement over the cooldown to approximately
535
degrees
F noted during the April 25,
1996 reactor trip discussed
in NRC
Inspection Report 50-400/96-05.
This improvement
was attributed to
better operator
performance in throttling AFW flow to the steam
generators.
Operators
noted indications of water
hammer in the turbine building
following the trip.
Plant personnel
performed walkdowns of the building
and found some minor damage to small pipe supports
and insulation.
The
inspector also conducted
an independent
walkdown and no additional
damage
was noted.
As noted in Section 01.2 above,
the licensee
conducted investigations
into the causes of the service water system problems.
The initial
engineering
assessment
of the "B" pump's shaft failure was documented in
the OHN-004 package.
An evaluation in accordance
with 10 CFR 50.59 was
performed for returning the plant to service with only the "A" NSW pump
available.
This evaluation concluded that while the
FSAR described the
plant as having two
NSW pumps, the
pumps were neither credited for
mitigating any Chapter
15 accidents
nor referenced
in any plant
Technical Specifications;
therefore,
having one unavailable for
maintenance
was acceptable.
The inspector
reviewed the
FSAR and
concluded that operating the plant with one
NSW pump available under
these
circumstances
was acceptable.
The OHH-004 package
was reviewed by the Plant Nuclear Safety Committee
(PNSC)
on September
7,
1996,
and permission to restart the plant was
granted
on September
9.
Investigations into the root causes for the "B"
NSW pump shaft break
and the "A" NSW pump initial trip were continuing
at the end of the inspection period.
c.
Conclusions
The post-trip review identified the direct cause,
evaluated
the
equipment actuations after the trip, and identified corrective actions
for equipment deficiencies.
01.4
Startu
Observations
a.
Ins ection Sco
e
71707
4
The inspector
observed the reactor
startup
on September
9,
1996 and
observed
generator
synchronization to the grid on September
10,
1996.
The inspector
observed
power ascension
to approximately
30 percent rated
. thermal
power .
b.
Observations
and Findin s
The inspector
observed that operators
were following procedures
and
ensuring that equipment
worked as expected.
Reactor criticality and
generator
synchronization
occur red with no problems.
Previous
problems
with a 6.9 kv unit auxiliary transformer breaker required that offsite
power
be fed through
a startup transformer breaker to the onsite 6.9 kv
distribution system.
Other secondary
equipment worked properly.
c.
Conclusions
The inspector
concluded that operator
performance during the startup
was
good.
04
Operator Knowledge and Performance
04. 1
Inadvertent Refuelin
Water
Stora
e Tank
Dilution Event
a.
Ins ection Sco
e
71707
90712
The inspectors
reviewed the circumstances
surrounding
an inadvertent
dilution of the
RWST.
In addition to gaining an understanding
of the
event, the inspectors verified that the licensee
per formed
a thorough
root cause determination
and that the proper
NRC reportability
determinations
were made.
b.
Observations
and Findin s
On September
6,
1996,
a licensed operator
was tasked with filling the
Condensate
Storage
Tank
(CST)
~
This direction was provided by the
shift's control
room supervisor.
The plant had been in Hot Standby
since September
3,
1996,
and the
CST was providing suction to the
system which was supplying the steam generator s.
The CST was already
filling using an alter nate source,
but was not filling at the desired
rate, thereby warranting the supplemental
action.
At approximately 7:30
p.m., the operator,
who had been
assigned. auxiliary operator duties that
shift, erroneously
proceeded to the
RWST pit and commenced fillingthat
tank by opening demineralized
water valve 1DW-5.
Approximately four
hours later, the
RWST HIGH LEVEL alarm was received in the main control
room.
Operators initially evaluated the alarm as being temperature
related.
Just'fter
midnight, following a routine check of RWST level
instruments
by the control
room operators,
they determined that the
was indeed filling.
The licensed auxiliary operator
was notified and,
realizing the error,
secured the
RWST fillingevolution.
The
RWST level increased
from 95 percent indicated level to 101 percent.
This equated to about 24,000 gallons of unborated
water
which, decreased
the boron concentration of the tank from 2476
ppm to approximately 2351
ppm.
Because this value was below the minimum limit of 2400
ppm
specified in Technical Specification Limiting Condition for Operation
(TS LCO) 3.5.4b, the
RWST was declared
The Boric Acid Tank
(BAT) had been declared
inoperable earlier because it was in a
recirculation lineup during an intentional fill process,
so the licensee
entered
TS
LCO 3.0.3 due to having the boration flow paths required by
TS
After securing the
RWST fill, plant
personnel
initiated actions to restore the boration systems to operable
status.
Normal standby
BAT alignment was restored
and its boration
flowpath declared operable within two hour s.
Operators
declared the
RWST oper able at approximately 2:00 p.m..on September
7, 1996, after
adding boric acid to the tank and mixing its contents with a containment
spray
pump operating in recirculation mode.
A 10 CFR 50.72 evaluation for immediate reportability concluded that the
reduced
during this event would have been
adequate to achieve the required
shutdown margin at all times of core
life.
This evaluation also concluded that the lower concentration
should comply with the TS Bases stipulation of maintaining
a pH value
between 8.5 and 11.0 for the water solution recirculated within
containment after
a design basis accident.
Based
on this information,
the event
was determined not to require immediate
NRC notification'as
a
condition outside of the plant's design basis.
However,
because of the
required entry into TS
the licensee
submitted
LER 96-20
(discussed
in report paragraph
08.4) in accordance
with 10 CFR 50.73.
The inspector discussed this event with plant personnel
and determined
that the communication to the operator
was clear in specifying the CST,
not the
RWST.
However, the operator
had developed
a mental picture of
the
RWST pit and was focused
on that tank when he was dispatched
by the
control
room.
The licensee's
investigation considered that
a
contributor could have been the failure to use specific valve numbers in
discussing
the upcoming task.
The inspector
found that filling the
was covered in section 8.0, Infrequent Operations,
of operating
procedure
Condensate
System,
Revision 7.
The inspector noted
that the operator did not have
a copy of the procedure
present.
This error (and the one discussed
below in report section 04.2)
represented
a continuation of the negative
human performance trend
discussed
in Inspection Report*50-400/96-07.
The significance of this
example
was that it caused the plant to be without an operable boration
flow path for several
hour s until system restoration
was accomplished.
Technical Specification 6.8.1.a
and Regulatory Guide 1.33, Revision 2,
Appendix A, Section 3.1 collectively require that written procedures
be
established.
implemented,
and maintained for energizing, filling,
venting,
and draining the auxiliary feedwater
system (for which the CST
is a safety-related
water supply).
Steps for fillingthe
CST [including
options to manipulate either
valve 1CE-23 (Step 2) or valves
1DW-490 and
1DW-486 (Step 3)3 are in subsection 8.3.2 of procedure
The
inadvertent
opening of valve lDW-5 and subsequent
dilution of the
resulted
from a failure to follow the steps in procedure
OP-134 and is
identified as
a violation of TS 6.8.1.a
(50-400/96-09-01).
Conclusions
The inadvertent
RWST dilution marked
a continuing adverse trend in human
performance
problems.
One violation was identified for failure to
properly implement operating procedures.
Inadvertent Actuation of Fuel Handlin
Buildin
Emer enc
Exhaust
Fan
Ins ection Sco
e
71707
The inspectors
reviewed the circumstances
surrounding
an inadvertent
actuation of a fuel handling building (FHB) emergency
exhaust
fan.
In
addition to gaining an understanding of the incident, the inspectors
verified that the licensee properly addressed
the root cause
and made
appropriate
recommendations
for corrective actions.
Observations
and Findin s
On September
29,
1996, auxiliary operators
were tasked with turning off
"spare" breakers
in various lighting and power panels throughout the
plant while conducting routine tours.
The operators
had been provided
an unapproved
and unverified list of breakers that was dated October
7,
1994 with the title "Spare Breakers
Found in ON Position".
While
performing the task,
one of the operators
turned off the breaker
for
circuit ¹17 in a power panel
labeled
PP 1-4833-SB.
This breaker
was on
the unverified list, but was part of the energized circuit powering
safety-related
radiation monitor RN-1FR-3564 B-SB.
With the operator's
action, the radiation monitor was de-ener gized which automatically
started
FHB emergency
exhaust
system fan, E-13B.
The fan started
as
designed
and caused
several
alarms in the main control
room, which was
the operators'irst
indication 'that
an error had occurred.
Upon
discovery,
the operators
stopped the evolution, generated
Condition
7
Report 96-03057,
and initiated actions to restore the radiation monitor
and secure the exhaust fan.
The inspector learned through reviewing the
CR and discussions
with licensee
personnel
that the operator
mistakenly
thought that all of the breakers identified on the list were spare
breakers
and that all were to be de-energized.
Because this error primarily involved performing changes to previously
completed
system lineups, the inspector
reviewed the licensee's
formal
process
as described in operations
management
manual
procedure
OHH-001,
Operations
- Conduct of Operations,
Revision 16.
Section 5.2.2 of this
procedure,
Electrical
and Valve Lineup Checklist, described the process
for documenting off-normal component positions, or positions not
specified
on component checklists in the various system operating
procedures.
Step 5.2.2.3.c directed personnel
to use Attachment
2 (of
OHH-001) to document the position of components for situations involving
small
changes to completed lineups;
The inspector
researched
the most
recently completed electrical lineup checklist for the radiation
monitoring system
(Attachment
1 to operating procedure
Radiation
Honitoring System,
Revision 4).
This review identified that the breaker
for circuit gl7 in panel
PP-1-4B33-SB
was last checked
and verified to
be
ON.
The inspector
concluded that had plant personnel
implemented the
instruction in OHH-001 step 5.2.2.3.c,
the completed proceduralized
lineup would have been
compared to the unreviewed list, which would have
identified the breaker's true function prior to the evolution taking
place.
The use of the unapproved list alone bypassed
the formal process
which the inspector concluded
was the root cause of the error .
Technical Specification 6.8.l.a
and Regulatory Guide 1.33, Revision 2,
Appendix A, Section 3.0 collectively require procedures f'r startup,
operation,
and shutdown of safety-related
PWR systems.
This requirement
is further implemented
by OHH-001, Section 5.2.1, Operating Unit
Procedure
Implementation,
which states in part that instructions for
energizing, filling, draining, starting up, shutting down,
and other
instructions appropriate for operations of systems
related to safety of
the plant shall
be delineated in system operating procedures.
The
licensee's
failure to implement procedure
OHH-001 step 5.2.2.3.c
contributed to the error resulting in the inadvertent actuation of FHB
emergency
exhaust
fan E-13B.
This failure to implement procedures
is
identified as the second
example of Violation 50-400/96-09-01
discussed
in section 04.1 above.
c.
Conclusions
Licensee
management
is continuing to address
the negative trend in human
performance in this area.
This trend was originally identified in NRC
Inspection Report 50-400/96-07.
A second
example to the violation
discussed
in section 04.1 of this report was identified f'r failure to
implement procedures.
07.1
a.
08.1
guality Assurance in Operations
Licensee
Self-Assessment
Activities
General
Comments
40500
During the inspection period, the inspectors
reviewed multiple licensee
sel f-assessment
activities, including:
~
Plant Nuclear Safety Committee
(PNSC) meetings
on September
5,
1996,
and September
25,
1996;
~
Nuclear Assessment
Section
(NAS) Audits on Steam Generator
Tube
Examination Records
(HNAS96-204),
and Operating
License
and
Technical Specification Compliance
(HNAS96-191).
These
assessments
were of good quality.
The assessment
documented in
HNAS96-191 lead to the identification of a missed
TS surveillance
requirement
(reference
report section
H8.7)
and demonstrated
good
questioning attitude on behalf of the
NAS auditor s.
Miscellaneous
Operations
Issues
(92700,
71707)
INPO Assessment
Ins ection Sco
e
71707
b.
The inspector
reviewed the
INPO report documenting
an assessment
that
was completed in Hay 1996.
This effort followed a preliminary review of
field notes discussed
in NRC Inspection Report 50-400/96-06.
Observations
and Findin s
The inspector
found that the issues identified were consistent with
those noted during the earlier review and with NRC perceptions of
licensee
performance.
No safety significant issues that required
immediate attention were identified.
C.
08.2
Conclusions
No regional
followup of the
INPO identified issues is planned.
0 en
LER 50-400/96-014:
Condition Outside of Design Basis in which
Two Charging/Safety Injection Pumps
(CSIPs)
were Inadver tently Connected
to the
Same
Emergency Electrical
Bus.
This
LER documented
the condition
described in NRC Inspection Report 50-400/96-06 in which the "B" CSIP.
and
"C" CSIP breakers
were both racked into the "B" 6.9kv electrical
bus
for approximately
10 minutes in 1989 during
a refueling outage.
The
licensee
rediscovered
the 1989 event while addressing
concerns
following
NRC identification of missing overload protective interlocks associated
with the emergency diesel
generators.
The 1989 event
happened with the
plant in Node 6 and no ongoing core alterations.
This event was
determined at that time not to be reportable
per
10 CFR 50.73 because
TS
action requirements for having no oper able diesels
(suspend
core
alterations
and vent the reactor)
wer e satisfied.
However, the
evaluation did not consider
the. situation
as
a condition outside of the
plant's design basis
(overloading the only operable diesel
generator).
Following the July 1996 NRC-identification of missing diesel
generator
overload protective interlocks (IR 50-400/96-06),
the inspectors
questioned
licensee
personnel
about the reportability of the 1989 event.
Per this review, the licensee
determined the condition to be reportable
on August 7,
1996 as
a condition outside of the plant design basis.
This
LER was issued
as
a result
and thoroughly described the
circumstances
concerning the 1989 event.
However,
completed corrective
actions
as stated in the
LER were limited to preventing recurrence of
the 1989 event (operating procedure revisions
and clearance tags).
Future corrective actions included plans to install mechanical
overload
protective interlocks for the charging
pump breakers
by the end of the
next refueling outage.
None of the corrective actions addressed
the
missed reportability call in 1989 for potentially placing the plant in
The
LER will remain open pending further
NRC
review of the licensee's
actions to address
the reportability aspect of
this issue.
08.3
0 en
LER 50-400/96-018:
Hanual
Reactor Trip due to Loss of Normal
This LER described the manual reactor trip that occurred
on September
3,
1996 following indications of a loss of the non-safety
related
normal service water
system (section 01.2 and 01.3).
Corrective
actions specified in the
LER were still ongoing at the end of the
inspection period and will be addressed
later.
This LER will remain
open pending further
NRC review of the licensee's
corrective actions.
08.4
0 en
LER 50-400/96-020:
Inadvertent
RWST Boron Dilution Event Caused
By Personnel
Error.
This
LER documented the event described in section
04.1 of this report.
The
LER properly described the circumstances
surrounding the inadvertent dilution and subsequent
TS
LCO 3.0.3 entry.
The safety significance of this event
was still under investigation
and
will be provided in a supplement to the
LER.
The
LER will remain open
pending receipt of the supplement
and
NRC review of the licensee's
corrective actions.
The
LER and its supplement will be tracked jointly
with the first example of Violation 50-400/96-09-01.
10
II. Maintenance
Hl
Conduct of Haintenance
H1.1
General
Comments
a.
Ins ection Sco
e
62707
The inspectors
observed all or portions of the following work
activities:
~
WR/JO 96-AEGY1, Replace
8 Normal Service Water Discharge Valve
~
WR/JO 96-AGBW1, Troubleshoot
8 Normal Service
Water
Pump
~
WR/JO 96-AJNN1,
8 Main Feed
Pump Repair
~
PH AMXF 001, Rebuild 8 Normal Service Water
Pump
b.
Observations
and Findin s
The inspectors
found the work performed under these activities to be
professional
and thorough.
All work observed
was performed with the
work package
present
and in active use.
Technicians
were experienced
and knowledgeable of their assigned
tasks.
The inspectors
frequently
observed
supervisors
and system engineers
monitoring job progress,
and
quality control personnel
were present
whenever required by procedure.
c.
Conclusions
Haintenance
personnel
were following procedures.
H2
Maintenance
and Material Condition of Facilities and Equipment
H2.1
Surveillance Observation
a.
Ins ection Sco
e
61726
The inspectors
observed all or portions of the following surveillance
tests:
HST-I0320, Train
8 Solid State Protection
System Actuation 5
Logic, Revision 12.
~
HST- I0070, Calibration of NIS Power Range
Overpower Trip High
Range Bistables,
Revision 3.
b.
Observations
and Findin s
The inspector
found that the testing
was adequately
performed.
c.
Conclusions
11
The surveillance
performances
were adequately
conducted.
H6
Haintenance
Organization
and Administration
H6. 1
On-line Maintenance
Plannin
62707
a.
Ins ection Sco
e
The inspector performed
a limited review of the licensee's
program for
scheduling
and performing on-line maintenance.
This program is
described in plant procedure
PLP-710,
Work Coordination Process,
Rev. 7,
WCW-001, Work Coordination
Manual Procedure,
Rev.
1,
and ADH-NGGC-0101,
Maintenance
Rule Program,
Rev. 3.
b.
Observations
and Findin s
These procedures
prescribe the responsibilities of management,
supervisory
and working level personnel
involved in the planning and
scheduling process.
Various desk top guides are provided to planners,
schedulers
and operators
in the work control center for .implementing the
program.
The work coordination process
makes
use of a 12-week rolling
window to schedule
work which has
been systematically planned.
Each
plant system is assigned
a specific week in the 12-week rolling window.
The weeks are either A train or
B train.'uring an A train week,
work
is not planned for 8 train equipment
and vice-versa.
The final schedule
for preventative or corrective on-line maintenance is approved at least
one week in advance.
The final schedule is approved
by both maintenance
and operations
supervision.
This schedule
al'lows plant personnel
to
prepare for work on specific safety systems in advance.
The preplanned rolling 12-week schedule
has
been reviewed
and approved
by the Probabilistic Safety Assessment
(PSA) Engineer to ensure that the
removal of systems
from service is acceptable
from a risk perspective.
If any combination not allowed by the preplanned matrix requires on-line
maintenance.
consultation with the
PSA engineer
is required before the
work can
be scheduled.
The
PSA engineer
is located conveniently at the
site.
Special
contingency plans or compensatory
measures
may be
appropriate
for risk-significant systems that are to be removed from
service.
c.
Conclusions
The inspector
concluded that the control of on-line maintenance
was
within the expectations
of NRC guidance.
Personnel
were aware of and
are trained
on the proper
and safe methodology to perform on-line
maintenance.
12
M8
M8.1
M8.2
M8.3
Miscellaneous
Maintenance
Issues
(92902,
90712)
General
Comments
90712
As evidenced
by the multiple LERs gener ated in recent months,
a trend
was noted in the number of recently identified procedural
deficiencies
that have,
in the past,
resulted in either missed Technical
Specification surveillance
requirements
or other problems associated
with equipment operability.
Several of the problems
have been
identified by the licensee's
ongoing review per Generic Letter 96-01,
Testing of Safety Related Logic Circuits.
Other
LERs involved unrelated
longstanding
procedural
deficiencies that affected various systems.
Many of the problems were historical in nature
and were identified as
the result of increased
licensee sensitivity to procedural
content
and
adherence
to Technical Specification requirements.
While many of the
specific deficiencies
have
been corrected,
several
corrective actions
were outstanding
and the
LERs remained
open.
These
LERs will be
reviewed individually and collectively with other
LERs for overall
surveillance
program adequacy.
A management
meeting in the
NRC Region
II office was planned for October 21,
1996 for licensee
management to
discuss
surveillance procedure
problems.
Discussions for each of the
LERs follows.
0 en
LER 50-400/96-002:
Failure to Properly Perform Technical
Specification Surveillance Testing.
Thi's
LER was discussed
in
Inspection Reports 50-400/96-02,
96-04, 96-05,
and 96-06,
and is the
result of the licensee's
ongoing Technical Specification Surveillance
Review related to Generic Letter 96-01.
Since Inspection Report
50-400/96-06
was issued,
three supplements to this
LER have been
generated.
Supplement
10 reported two additional deficiencies.
One
involved load sequence
timing circuits for the Emergency Safeguards
Sequencer.
The other dealt with parallel circuits which operate
an
inlet damper
associated
with computer/communication
room emergency
ventilation.
Supplement
11 reported three additional testing
deficiencies
associated
with ventilation systems for the
ESW intake
structure,
the
125
VDC emergency battery rooms,
and the control
room.
Supplement
12 discussed
inadequate testing of the respective
pressure
control valves for each motor-driven
AFW pump.
This
LER and its
supplements will remain open until further
NRC review of the licensee's
corrective actions,
including retest
requirements,
is completed.
0 en
LER 50-400/96-010:
Residual
Heat Removal
(RHR) System
Surveillance Testing Deficiency that Caused
Past Entries Into TS 3.0.3.
This
LER described
procedural
revisions (to test procedures
OST-1008,
1A-SA RHR Pump Operability Quarter ly Interval
Modes 1-2-3;
and OST-1092,
18-SB
RHR Pump Operability Quarter ly Interval
Modes 1-2-3) in October
1992 that resulted in several
subsequent
entries into TS
during
RHR system testing.
The situations
arose
from valve alignments
which cross-tied the two redundant trains of RHR while one was
inoperable for the test.
The proceduralized
valve line-ups resulted in
the operable train being unable to provide the minimum required low head
13
safety injection flow to the
RCS during
a postulated large break loss of
coolant accident.
Supplement
1 to this
LER was issued in July 1996 to describe
a similar
situation that had been discovered while investigating the
RHR scenario.
A deficiency in original procedure
HST-I0417, Containment Ventilation
Isolation Area Radiation Honitors Relay Actuation Logic Test, resulted
in both redundant
containment
vacuum relief system valves isolating.
This caused
both trains of the containment
vacuum relief system to be
inoperable for approximately 45 minutes during each performance of the
test since
1987.
This item had been originally identified as
a concern
due to required entries into TS 3.0.3 back in December
1995, but the
repor tability aspect of such entries
was missed.
This LER supplement
addressed
corrective actions for both the procedural
deficiencies
and
the missed reportability determination in December.
Supplement
2 was issued during this inspection period and discussed
the
safety significance of both of the above items.
The licensee's
analysis
of the
RHR situation demonstrated
that while the
pump not being tested
would have inadequately split flow between
a recirculation path cr eated
by the test lineup and the intended
RCS path, the availability of the
pump under test would have also provided split flow through
a
recirculation path
and the
RCS.
The combined flows from each of the
pumps to the
RCS injection path were determined
by the licensee to meet
minimum system flow requirements for accident conditions.
A
probabilistic safety assessment
was also 'cited in the
LER as having
analyzed the degraded testing condition and determining the increase in
annual
core
damage
frequency to be 0.0065 percent.
An analysis of the
containment
vacuum relief system situation found that reliance
on
operator
action would be needed to mitigate the effects of an
inadvertent actuation of the containment
spray system (for which the
system
was designed)
coincident with the test lineup.
This
LER will remain open pending
an independent
NRC review of the above
items, their safety consequences,
and the licensee's
corrective actions.
Any enforcement
decisions will be made following the additional review.
H8.4
0 en
LER 50-400/96-011:
Inadequate
Surveillance
Procedures
Failed to
Provide
a Heans for Identifying Deactivated Automatic Containment
Isolation Valves Which Are to be Subjected to Verification Every 31 Days
in Accordance with Technical Specifications.
Two automatic containment
isolation valves,
had been deactivated
and shut in
December
1995 after
1FW-221 failed stroke time testing required by
During a walkdown in June
1996,
an operator questioned
whether
or not the deactivated
shut status of the valves
had been
verified every 31 days
as required by TS Surveillance Requirement 4.6.1.1.a.
The licensee
determined later that compliance with
TS 4.6. 1. l.a had not been met for these valves.
The cause of the
failure to perform the required verification was
an incorrect
interpretation of the TS 4.6.1.1.a
requirement during initial procedure
development.
The earlier interpr'etation did not consider those
penetr ations which were normally capable of automatic isolation, but
14
H8.5
H8.6
whose isolation valves were deactivated
for compliance with other
Technical Specifications.
The safety consequences
of the failure to perform the monthly
verification were mitigated by the fact that the deactivated
valves were
placed under clearance
in accordance
with plant procedures.
Equipment
under
clearance
is usually reviewed by the Superintendent
of Shift
Operations at each shift turnover.
The two valves in question
were
verified to be deactivated
and shut on June 30,
1996.
Corrective
actions to revise procedures
and perform an additional
review for
possible similar situations with other containment isolation valves were
still ongoing at the end of the inspection.
The
LER will remain open
pending licensee
completion of the corrective actions
and further
NRC
review.
0 en
LER 50-400/96-015:
Unplanned Partial
Engineered Safety Feature
Actuation (Hain Steam Isolation Signal) During Surveillance Testing
Due
to Operator Error .
This
LER described
the event discussed
in detail in
NRC Inspection Report 50-400/96-07 which was example
2 of Violation 96-
07-01.
The corrective actions stated in the
LER, along with additional
corrective actions to address
the recent negative trend in personnel
error s, will be tracked in conjunction with the violation.
This LER and
the violation will remain open pending licensee
completion
and
NRC
review of corrective actions.
0 en
LER 50-400/96-016:
Failure to Perform Reactor Trip Bypass
Breaker
Surveillance
Testing Required by Technical Specifications.
This
LER reported
a Technical Specification violation in which the licensee
failed to test reactor trip bypass
breakers prior to placing them in
service.
This was
a historical
problem that dated
back to original
development of the surveillance
procedures
(HST-I0001, Train A Solid
State Protection
System Actuation Logic & Haster
Relay Test;
and
HST-I0320, Train
B Solid State Protection
System Actuation Logic &
Haster Relay Test) in 1986.
The licensee attributed the procedural
oversight to a possible misinterpretation of the term "in-service" and
how it applied to the reactor trip bypass breakers.
A licensee
review
of the
FSAR later found that it contained conflicting information with
the TS related to testing the breakers.
The
FSAR incorrectly stated
that the licensee
"does not propose to verify the operability of each
bypass
breaker prior to placing it in service
each time the main reactor
trip breakers
are to be tested."
The safety consequences
of not testing the bypass
breakers prior to
placing them in service
was mitigated by the fact that the
breakers'emote
manual
shunt trips were tested
by the
same surveillance
procedures
after placing them in service but prior to removing the main
reactor trip breakers
from service.
Corrective actions will include
revising the
HSTs and conflicting section of the
FSAR,
and training
procedure writers to emphasize
the importance of fully under standing
TS
testing requirements.
The
LER will remain open pending licensee
completion
and
NRC review of the corrective actions.
15
H8.7
0 en
LER 50-400/96-017:
Failure to Perform Surveillance Testing
Required by Technical Specifications.
This
LER described
a missed
18-
month surveillance test for the Reactor Auxiliary Building Emergency
Exhaust
System.
The test
was scheduled to be performed during Refueling
Outage
Number Six in October
1995.
A recent Nuclear Assessment
Section
audit of the plant's
document control program could not find a copy of
the completed procedure in the records storage vault.
The procedure
(OST-1052,
Reactor Auxiliary Building Emergency Exhaust
System,
18-month
Operations Surveillance Test) did not turn up during subsequent
searches
of other offices.
A review of control
room logs indicated that the test
had not been performed
even though the licensee's
tracking system for
Technical Specification surveillance tests indicated successful
test
performance
on October 5, 1995.
The 18-month frequency established
by
the TS for. this test would have
been exceeded
on Harch 11,
1996 based
on
the last previous completion in April of 1994'he
emergency
exhaust
system
was successfully tested
on June 6,
1996 (prior to the
identification of this issue
and done to resolve another
issue
referenced
in LER 96-009).
Failure to perform the test between
Harch
and June of 1996 constituted
a violation of TS 4.7.7.d.
The licensee
had not completed its investigation into the cause of this
event
and its safety consequences.
The results of that investigation,
along with recommended
corrective actions, will be presented
in a
supplement to the LER.
The
LER will remain open pending issuance of the
supplement
and further
NRC review.
Possible
enforcement
actions will be
considered
at that time.
The inspectors
considered
the identification
of this issue
by the licensee's
NAS organization to be an example of
good performance.
H8.8
Closed
Unresolved
Item 50-400/96-07-02:
Corrective Actions for
Haintenance
Rule Implementation
Problems
This item was unresolved
pending licensee
completion
and
NRC review of
an impact assessment f'r maintenance
rule scoping problems.
Inspection
report 50-400/96-07
documented that
NRC inspectors
found
a problem with
the maintenance
rule scoping for Boric Acid Filter Inlet Valve 1CS-559
in the Equipment Data Base
System
(EDBS).
Valve 1CS-559
was designated
in EDBS as not within the scope of the maintenance
rule,
even though it
is in the emergency boration flow path which is used to mitigate
accidents.
The licensee informally decided to use this existing data
base to implement the expert panel
maintenance
rule scoping
determinations.
The expert panel
scoping determination
was accurate in
identifying the boration flow path (which includes
as
a
maintenance
rule function.
However, the valve had been identified in
the wrong system (filter backwash)
when the
EDBS was originally created.
The filter backwash
system
had been determined
by the expert panel
not
to have
a maintenance
rule function and all of its components
were
classified in the
EDBS as Haintenance
Rule
- No.
The erroneous
classification of 1CS-559 in the
EDBS,
and other errors discovered later
by the licensee,
indicated that the licensee
had not done
an adequate
quality check of the
EDBS prior to implementation of the maintenance
rule.
16
The inspector
reviewed the licensee's
root cause investigation for
condition report. 96-2175, written to address
the
NRC finding.
The
licensee
reviewed approximately
1200 components that could have
been
subject to the
same misclassification
as
and found multiple
components
in each of 9 different systems, that were not scoped correctly
in EOBS because
they wer e listed in incorrect systems.
In addition, the
licensee
found boundary valve scoping problems, similar to that of 1CS-
559, with multiple components
in 3 systems that interface with the
chemical
and volume control
system requiring maintenance
rule scoping
changes.
The licensee
performed
an impact assessment,
covering the last
3 years, to determine if any failures of maintenance
rule components
had
been missed
and concluded that one
had been missed,
but it was
determined
not
a maintenance
preventable
functional failure.
The root
cause investigation also determined that
an action item previously
assigned to review this exact area (in early 1996)
had been left off the
site wide tracking system
and was
a significant contributor to this
problem.
Inspection report 50-400/96-07
had noted previous
opportunities to fully identify and fix this problem prior to the
July 10,
1996 implementation of the maintenance
rule.
10 CFR 50.65 (b) establishes
that the scope of the maintenance
rule
monitoring program shall include safety-related
structures.
systems,
or
components that are relied upon to remain functional during and
following design basis events to ensure the integrity of the reactor
coolant pressure
boundary,
the capability to shut
down the reactor
and
maintain it in a safe shutdown condition,. and the capability to prevent
or mitigate the consequences
of accidents that could result in potential
offsite exposure
comparable to the
10
CFR part 100 guidelines;
and non-
safety related structures,
systems or components that are relied upon to
mitigate accidents or transients
or are used in the plant emergency
operating procedures,
or whose failure could prevent safety-related
structures,
systems,
and components
from fulfillingtheir safety-related
function, or whose failure could cause
a reactor
scr am or actuation of a
safety-related
system.
Procedure
ADA-NGGC-0101, Haintenance
Rule Program,
Revision 3,
implements
10 CFR 50.65 and provides maintenance
rule implementation
instructions.
Section 9.3.1,
under scoping, directed personnel
to
obtain systems lists from the Equipment Data Base System
(EDBS) and
supply to the Expert Panel
for evaluation.
Attachment 1, List of CP&L
Haintenance
Rule Systems, lists the
EDBS system
name,
system
number,
and
the exper t panel determination.
The expert panel
system determinations
were loaded into EDBS with all components in each
system receiving the
determination
for the system.
The inspectors
concluded that the improper
scoping of 1CS-559
and the
other
components
found by the licensee
was
a violation of 10 CFR 50.65(b)
under the maintenance
rule (50-400/96-09-02).
17
III. En ineerin
E1
Conduct of Engineering
El.l
General
Comments
37551
Engineering support for various activities including the
NSW pump
troubleshooting effort and routine support during the September
9
reactor startup
was good.
E1.2
Technical
S ecification Inter retation 95-003
a.
Ins ection Sco
e
37551
The inspector
reviewed Technical Specification Interpretation
(TSI) 95-
003, Ultimate Heat Sink (UHS), Revision 3,
and referenced
documentation
to determine
why the interpretation
was needed,
what affect it had on
the associated
Technical Specifications
(TS) and Final Safety Analysis
Report
(FSAR),
and whether regulatory requirements
were met.
b.
Observations
and Findin s
The TSI stated that TS 3.7.5 Limiting Condition for Operation
(LCO)
action statement
should be applied for the main reservoir
at the 215
foot elevation (instead of the TS 205.7 foot elevation value);
and apply
the action at 94 degrees
F for either reservoir with at least
one
emergency
pump running, or use 86 degrees
F if neither
emergency service water
pump is running (instead of the TS reservoir
temper ature value of 95 degrees
F).
TSI 95-003, Revision
0 was issued
on June 28,
1995 and included only the
215 foot elevation for applying the TS 3.7.5
LCO action.
The TSI stated
that the limit is expected to be temporary,
and will remain in effect
until engineering analysis
supports rescission.
Revision
1 was issued
September
1,
1995, which included the 215 foot elevation
and established
a new reservoir
temperature limit of 89 degrees
F to apply the
LCO
action.
Revision
2 was issued October 5, 1995, which included the 215
foot elevation,
but changed temperature to apply the
LCO action at 95
degrees
F for either
reservoir with at least
one emergency service water
pump running, or use 86 degrees
F if neither
emergency service water
pump is running.
Revision 3 was issued
Harch 29,
1996, which included
the 215 foot elevation,
but lowered the 95 degrees
F to 94 degrees
F,
leaving the 86 degrees
F as previously stated.
The TSI and revisions
were reviewed
and approved
by plant management
and the Plant Nuclear
Safety Committee.
TSI Basis
The inspector
researched
the basis for this TSI through review of
Engineering Service Requests
(ESRs)
9500548,
9500707,
9500726.
Adverse
Condition and Feedback
Reports
(ACFRs) 95-00129
and 95-01464,
18
and 3.7.4,
and the associated
TS bases.
The ACFRs and
ESRs were
generated
as
a result of question
number
8 from the Service Water System
Operational
Performance
Inspection
(SWSOPI)
conducted
by the licensee.
The SWSOPI questioned
how the TS 3.7.5 value of 95 degrees
F would
prevent the ultimate heat sink from exceeding the 30 day post accident
value of 95 degrees
F.
A calculation effort to determine the
appropriate starting value was what resulted in the 94/86 degree
F
reservoir
temperature
value.
The licensee
discovered during the temper ature resolution effort that
although the TS 3.7.5
LCO main
r eservoir
level of 205.7 feet was
adequate
volume to dissipate
the design heat load over
a 30 day period
(stated in TS Bases),
the level
was not adequate for the emergency
system to provide sufficient flow for all components to
remove the design heat generated.
Since
an
ESW pump upgrade could not
be immediately conducted to incr ease flow, an alternative that could be
immediately put in place
was to raise the limit on reservoir level.
A
calculated reservoir level of 215 feet was determined to provide
adequate
flow to all components
serviced
by ESW to maintain them
This administrative limit was put in place through the TSI.
The inspector's
review of PNSC meeting minutes for the TSI found that
the licensee felt the TSI required
them to apply the
LCO "early" as
compared to the TS value and therefore the interpretation
was more
conservative
than TS.
No discussion of a license
amendment/TS
change
was documented.
The inspector
found that the conclusions of ESR 9500548,
issued August
25.
1995,
were that the 215 foot elevation main reservoir administrative
limit (TSI) needed to be maintained to ensure operability of some
components.
The inspector
found that later
ESRs used the 215 foot
elevation
as
an assumption
and did not support retr acting that limit.
The
FSAR change
markup did not include the 215 foot elevation
administrative limit and stated that
a license
amendment
was not
necessary.
The
ESR stated that main reservoir level
has never
been
below approximately 217 feet elevation during the history of the plant.
The inspector
reviewed
FSAR section 2.4.11.7,
Heat Sink Dependability
Requirements,
FSAR section 9.2.1, Service Water System,
and
FSAR section
9.2.5, Ultimate Heat Sink.
None of the
FSAR sections
mention the main
reservoir level needing to be 215 feet to provide adequate
ESW flow for
heat removal.
These sections
were changed
by Amendment 46, submitted
April 12,
1996, to include many of the results
from the
SWSOPI question
on this issue.
ESR 9500707,
issued January
17,
1996, provided the basis
for the
FSAR changes to those sections
and included those
from ESR
9500548,
but made
no mention of the 215 foot level.
The 10 CFR 50 '9
safety evaluation for ESR 9500707 concluded that
a license
amendment
was
needed for UHS temperature.
No change
was mentioned for level.
Licensee Administrative Procedure
(AP) 107, Technical Specification
Interpretations,
Revision 7. governs the TSI process
and indicated that
TSIs are intended to provide
a consistent
and reasonable
position on the
meaning
and applicability of the TSs.
An annual
TSI audit required by
19
AP-107 was conducted
and issued
on December
29,
1995.
It concluded that
TSI 95-003 should remain in place
and did not need to be incorporated
into a TS change.
The reason
given in the audit report for
a TS change
not being needed
was that it was "too costly."
The inspector noted that
TSIs are not required by AP-107 to receive
a review under
Licensee
management
had authorized the regulatory affair s organization
to develop
a TS change
on May 21,
1996 that encompassed
TSI 95-003.
The
licensee
had not made
any meaningful
progress
on the change prior to the
inspector
discussing
the issue.
Licensee Corrective Actions
The inspector
found that the licensee
was adequately
pursuing corrective
action of the technical
issue
from a hardware standpoint including
modifications during the upcoming outage
(RF07) to replace the "B" ESW
pump with a larger capacity
pump and to enlarge the containment
fan
cooler orifices to allow greater flow.
The replacement of the "A" pump
is scheduled for the following outage
(RF08).
As a result of the
inspector's
concern regarding whether the TS met the requirements of 10 CFR 50.36, the licensee
intends to submit
a TS change
by November 1,
1996,
and revise the
FSAR to include all the information in the TSI.
However, at the time of the inspection the TSI remained the only safety
barrier.
c.
Conclusions
Licensee corrective action for dealing with the hardware
and design
problem was prompt.
However, this issue is unresolved
based
on the
inspector reviewing the previous main reservoir level history to
determine whether the plant operated outside the design basis,
and an
overall regulatory impact assessment
including whether
the TS and
met regulatory requirements
(Unresolved Item 50-400/96-09-03.
UHS/ESW
Technical Specification Interpretation).
E7
Quality Assurance in Engineering Activities
E7.1
S ecial
FSAR Review
37551
A recent discovery of a licensee operating their facility in a manner
contrary to the Updated Final Safety Analysis Report
(UFSAR) description
highlighted the need f'r a special
focused review that compares plant
practices,
procedures
and/or parameters
to the
FSAR descriptions.
While
performing the inspections
discussed
in this report, the inspectors
reviewed the applicable portions of the
FSAR that related to the areas
inspected.
The licensee
made
a presentation
to the
NRC on May 31,
1996 concerning
their corporate-wide
plan for reviewing the
FSAR at the
CPSL sites.
The
program has generated
a large
number of condition reports at the Harris
Plant
(252 by the end of the inspection period).
Three
have resulted in
LERs.
The results
from this program will be reviewed in the closure of
20
Unresolved
Item 50-400/96-04-04,
Tracking
FSAR Discrepancy Resolution.
A discrepancy
was identified by the inspectors
between the
FSAR wording
and plant engineering
documents
related to the emergency service water
system
and ultimate heat sink as described in section El.2.
E8
E8. 1
Hiscellaneous
Engineering Issues
(92700)
0 en
LER 50-400/96-013:
Condition Outside of Design Basis where the
RWST Had Been Connected to
a Non-seismically Qualified System.
This
LER
reported
a long-standing
design
and operational
issue related to non-
seismic portions of piping systems
being connected to the Refueling
Water Storage
Tank
(RWST) during routine operations.
Specifically, the
non-seismic portion of the spent fuel pool purification system
had
routinely been aligned to the
RWST for pool cleanup,
and non-seismic
piping to the hydrostatic test
pump had been aligned to the
RWST for
fi'lling the safety injection accumulator
tanks inside containment.
These problems dated back to original procedural
development
and the
failure to adequately reconcile operating procedures
with the plant's
design basis.
The licensee
evaluated the safety significance of not having the
available during
a seismic event using probabilistic risk assessment
and
determined that the increase to conditional core
damage
was small.
Currently, valves isolating the non-seismic portion of the systems
trom
the
RWST are locked shut
and under clearance.
The inspectors
have
verified these valve configurations during routine plant walkdowns.
The
licensee
was still developing contingency plans for purifying the spent
fuel pools
and fillingthe safety injection accumulators
at the end of
the inspection period.
This LER will remain open pending completion of
the licensee's
ongoing actions
and further review by the inspectors.
R1
R1.1
P2
P2.1
IV. Plant
Su
rt
Radiological Protection
and Chemistry (RPK) Controls
General
Comments
71750
The inspector
observed radiological controls during the conduct of tours
and observation of maintenance activities and found them to be
acceptable.
The general
approach to the control of contamination
and
dose for the site was good.
Teamwork between the various departments
continued to be
a major contributor to the good control of dose,
A
resin spill in the waste processing building that occurred during this
period was handled well.
Status of Emergency Preparedness
(EP) Facilities, Equipment,
and
Resources
Hurricane
Fran
71750
71707
21
a.
Ins ection Sco
e
The licensee
made preparations for Hurricane Fran prior to and on the
September
5,
1996, arrival of the hurricane.
Licensee
Procedure
AP-301,
Adverse Weather Operations,
Revision 15,
was used for the preparations.
The inspectors
observed
the licensee's
preparations
and response
to
determine whether they met regulatory requirements.
b.
Observations
and Findin s
The Harris plant was shutdown in Hot Standby
due to the loss of normal
service water on September
3,
1996,
as discussed
in section 01.2 of this
report.
The Hurricane caused
no damage to safety or balance of plant
equipment
on site.
The only damage
observed
on site were several
panels
from the administration building that blew off and were found after the
hurricane against the security fence.
Offsite equipment
such
as evacuation
sirens lost power during the
hurricane.
Two reports were made under
10 CFR 50.72 by the licensee
related to offsite situations.
The first was on September
6,
1996 when
the licensee
began losing power to the offsite emergency evacuation
The report was
made when 20 percent of the sirens
were lost.
When the hurricane
was over the licensee
had lost power to approximately
90 percent of the 81 offsite sirens.
The local area
had received
significant rain for the week prior to the arrival of the hurricane
which had softened the ground.
As a result,
a significant number of
trees
were uprooted by the Hurricane winds which gusted to 80 miles per
hour.
These trees fell on the power lines which caused the loss of
power
.
The licensee
had returned
90 percent of the sirens to service by
September
9,
1996 when the
NRC (and FEN) gave permission for the plant
to restart.
The second report related to restricted plant access
due to downed trees
and power lines on the plant access
road and local area roads.
Sufficient additional operators
were available to maintain minimum
required shift coverage.
The 10 CFR 50.72 report was
made at 5:41 a.m.
on September
6, 1996.
Access to the plant was restored
by approximately
9:00 a.m. that
same day.
LER 96-019 addresses
this issue.
c.
Conclusions
The licensee appropriately
responded to the hurricane
and made the
proper notifications as required by the regulations
when emergency
response
capability was hampered
and emergency
equipment
was
unavailable.
P8
Hiscellaneous
EP Issues
(71750)
P8.1
Closed
LER 50-400/96-019:
Hurricane Fran Hampered Site Personnel
From
Accessing
The Plant.
This LER reported
degraded
road conditions which
were caused
by the hurricane
on September
5,
1996,
and were determined
by the licensee to have significantly hampered site personnel
in the
performance of their duties.
Approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the initial
22
S1
S1.1
S1.2
a.
b.
1-hour non-emergency notification was made to the
NRC, plant access
roads were cleared which allowed site personnel
to get into and out of
the plant.
No further corrective actions
were required for this
condition and this
LER is closed.
Further discussion of the hurricane
appears
in section
P2.1 of this report.
Conduct of Security and Safeguards Activities
General
Comments
71750
The inspector observed security
and safeguards activities during routine
plant tours
and during the response to Hurricane Fran on September
5 and
6, 1996.
Compensatory
measures
during the hurricane,
including security
staff support to plant personnel,
were good.
General
Inte
r it of Protected
Area
Barriers
Ins ection Sco
e
71750
The inspectors
toured the facility to ensure that protected
area
and
vital area barriers
were separated
and that the barriers
had no openings
and. were not damaged or in an otherwise degraded
condition.
Observations
and Findin s
During a plant tour
on September
13,
1996, the inspector
observed
protected
area gate
1B open approximately three to four feet.
This gate
is located at the vehicle "tr ap" adjacent to the security building and
is the protected
area gate where vehicle access is granted.
The outer
vehicle trap gate
was secured.
The inspector
was concerned
because
there
was
an opening into the protected
area
and security personnel
were
neither
aware of the opening nor in a position to 'fully observe
and
control access
into the protected
area.
The inspector
noted that the
security personnel
were located in a gate
house inside the vehicle trap
(approximately
20 feet away)
and appeared to be engaged in duties
related to a vehicle that was exiting the plant.
The inspector gained
the attention of the officer in the gate
house
who approached
the gate.
The inspector informed the officer that the gate
was open
and the
officer, via radio communications with the Facility Access Control
(FAC)
operator,
directed complete closure of the gate.
The inspector reported this finding to security management
personnel
who
initiated Condition Report 96-02886
and an investigation.
The
licensee's
investigation determined that the gate
had only been
opened
for approximately two minutes
and that the officer failed to ensure
complete closure of the gate once the vehicle was inside the trap.
The
officer had directed the
FAC to close the gate
once the vehicle was
inside and then turned his back on the gate in route to the gate house.
The
FAC operator
apparently did not complete the gate closure
sequence.
Once the officer was informed by the inspector that the gate
was opened,
prompt actions to secure the gate were taken.
23
The inspector determined through discussions
with licensee
personnel
and
reviews of security procedures
that the officer failed to perform
specified duties
as outlined in procedures
Security Search
and
Contraband Denial, Revision 5;
and SP-004,
Security Post Duties,
Responsibilities,
and Patrol
Procedures,
Revision 2.
Procedure
Section
12 delineates
the duties of guards/armed
responders
which
includes ensuring integrity of protected
area barriers.
Procedure
Section 7.0, Vehicle Searches,
Step
7 specified that, following
access
by the vehicle, the gate
and the vehicle barrier
system active
barrier shall
be closed
and returned to the blocking position.
These
procedures
implement the licensee's
security plan
(CP8L Physical
Security and Safeguards
Contingency Plan,
Revision 0).
Section 4.0 of
this plan prohibits protected
area barrier
openings of greater
than 96
square
inches.
The failure to secure the protected
area barrier created
a short-term vulnerability in that the integrity of the protected
area
bar rier was not maintained
and
an opening in excess of the requirements
in the security plan existed for approximately two minutes.
The
inspector investigated the vehicle trap area
by observing gate
operations
from the
FAC and locally surveying accessibility to the
vulnerable area.
The inspector
concluded that although the officer in
the gate
house
and personnel
at three other watch stations with video
monitoring capability were unaware that the gate
was left open, the
vulnerability was neither easily or likely to be exploitable.
Technical Specification 6.8.1.c requires that written procedures
shall
be established,
implemented,
and maintained covering Security Plan
implementation.
The failure to properly secure
gate
1B constituted
a
failure to follow procedure
and is a Violation of TS 6.8.1.c
(50-
400/96-09-04).
Immediate corrective actions for this error included counseling the
officer and training other security personnel
on this event
and the
contributing factors.
The inspector verified that the incident was
recorded in the Quarterly Safeguards
Event Log in accordance
with the
requirements of 10 CFR 73.71, Reporting of Safeguards
Events.
Security
management's
response to this finding was prompt and appropriate.
c.
Conclusions
The protected
area
boundary was generally intact;
however, the inspector
identified a short-term vulnerability caused
by personnel
error.
One
violation was identified for failure to follow security procedures.
F1
Control of Fire Protection Activities
F1.1
General
Comments
71750
The inspector
observed fire protection equipment
and activities during
the conduct of tours
and observation of maintenance activities and found
them to be acceptable.
24
X1
Exit Meeting Summary
V. Mana ement Meetin s
The inspectors
presented
the inspection results to members of licensee
management
at the conclusion of the inspection
on October
18,
1996.
The
licensee
acknowledged the findings presented.
The inspectors
asked the licensee
whether
any of the material
examined
during the inspection should
be considered proprietary.
No proprietary
information was identified.
25
Licensee
PARTIAL LIST OF PERSONS
CONTACTED
D. Alexander,
Super visor, Licensing and Regulatory Programs
D. Batton, Superintendent,
On-Line Scheduling
D.
Br aund,
Superintendent,
Security
B. Clark, General
Hanager,
Harris Plant
A. Cockerill, Superintendent,
I8C Electrical
Systems
J. Collins, Hanager,
Training
J.
Dobbs,
Hanager,
Outage
and Scheduling
J.
Donahue,
Director Site Operations,
Harris Plant
R. Duncan,
Superintendent.
Hechanical
Systems
W. Gautier,
Hanager,
Haintenance
W. Gurganious,
Superintendent,
Environmental
and Chemistry
H. Hamby,
Super visor, Regulatory Compliance
H. Hill, Hanager,
Nuclear Assessment
D. HcCarthy, Superintendent.
Outage
Hanagement
K. Neuschaefer.
Superintendent,
Radiation Protection
W. Peavyhouse,
Superintendent,
Design Control
W. Robinson,
Vice President,
Harris Plant
G. Rolfson.
Hanager,
Harris Engineering Support Services
S. Sewell,
Hanager,
Operations
T. Walt, Hanager,
Performance
Evaluation
and Regulatory Affairs
NRC
T. Le, Harris Project Hanager,
H. Shymlock, Chief, Reactor Projects
Branch 4
26
IP 37551:
IP 40500:
IP 61726:
IP 62707:
IP 71707'P
71750'P
90712:
IP 92902:
IP 93702:
INSPECTION PROCEDURES
USED
Onsite Engineering
Effectiveness of Licensee Controls in Identifying, Resolving,
and
Preventing
Problems
Surveillance Observations
Maintenance Observation
Plant Operations
Plant Support Activities
In-office Review of LERs
Followup
- Maintenance
Onsite Response to Events
~0ened
50-400/96-09-01
50-400/96-09-02
50-400/96-09-03
50-400/96-09-04
Closed
ITEMS OPENED,
CLOSED.
AND DISCUSSED
Failure To Follow Operating Procedures,
Paragraph
04.1.
Inadequate
Maintenance
Rule Scoping,
Paragraph
M8.8.
Technical Specification Interpretation for Ultimate
Heat Sink 5 FSAR Discrepancy,
Paragraph
E1.2
Failur e To Properly Secure Protective Area Gate,
Violating Security Procedures,
Paragraph
S1.2.
50-400/96-07-02
Corrective Actions for Maintenance
Rule Implementation
Problems,
Paragraph
M8.8
50-400/96-019
LER
Hurricane Fran Hampered Site Personnel
from Accessing
the Plant,
Paragraph
P8.1.
Discussed
50-400/96-014
LER
50-400/96-018
LER
50-400/96-020
LER
50-400/96-002
LER
50-400/96-010
LER
50-400/96-011
LER
Condition Outside of Design Basis in which Two
Charging/Safety Injection Pumps
(CSIPs)
were
Inadvertently Connected to the
Same
Emergency
Electrical
Bus,
Paragraph
08.2.
Manual Reactor Trip due to Loss of Normal Service
Water,
Paragraph
08.3.
Inadvertent
RWST Boron Dilution Event Caused
by
Personnel
Error, Paragraph
08.4.
Failure to Proper ly Perform Technical Specification
Surveillance Testing,
Paragraph
M8.2.
Residual
Heat Removal
(RHR) System Surveillance
Testing Deficiency that Caused
Past Entries into TS
3'.3,
Paragraph
M8.3.
Inadequate
Surveillance
Procedures
Failed to Provide
a
Means for Identifying Deactivated Automatic
Containment Isolation Valves which are to be Subjected
50-400/96-015
LER
50-400/96-017
LER
50-400/96-013
LER
50-400/96-016
LER
27
to Verification every 31 Days in Accordance with
Technical Specifications,
Paragraph
H8.4.
Unplanned Partial
Engineered Safety Feature Actuation
(Hain Steam Isolation Signal) During Surveillance
Testing
Due to Operator. Error, Paragraph
H8.5.
Failure to Perform Reactor Trip Bypass Breaker
Surveillance Testing Required by Technical
Specifications,
Paragraph
H8.6.
Failure to Perform Surveillance Testing Required by
Technical Specifications,
Paragraph
H8.7.
Condition Outside of Design Basis where the
RWST had
been Connected to a Non-seismically Qualified System,
Paragraph
E8.1