ML18010A813
| ML18010A813 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/29/1992 |
| From: | Christensen H, Shannon M, Tedrow J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18010A811 | List: |
| References | |
| 50-400-92-17, NUDOCS 9210140151 | |
| Download: ML18010A813 (22) | |
See also: IR 05000400/1992017
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I I
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report No.:
50-400/92-17
Licensee:
Carolina
Power
and Light Company
P. 0.
Box 1551
Raleigh,
NC 27602
Docket No.:
50-400
Licensee
No.:
Facility Name:
Harris
1
)
Inspection
Conducted:
August
22 - September
25,
1992
Inspectors:
J.
e r w
Senior Resident
Inspector
V 2'p FM-
Da e
S gned
M.
Sh
no
,
Re 'dent Inspector
Approved by:
H. Christensen,
Section Chief
Di.vision of Reactor Projects
Da e
S gned
Per ~~
ate Signed
SUMMARY
Scope:
This routine inspection,was
conducted
by two resident
inspectors
in the areas
of plant operations,
radiol'ogical controls, security, fire protection,
surveillance
observation,
maintenance
observation,
reliable decay heat
removal
during outages,
annual
emergency drill, and licensee
event reports.
Numerous
facility tours were conducted
and facility operations
observed.
Some of these
tours
and observations
were conducted
on backshifts.
Results:
Three violations were identified:
An apparent violation was identified regarding the failure to promptly
identify and correct
an adverse
condition involving the charging/safety
injection alternate mini-flow system,
paragraph
7.a.
A violation was also identified regarding the failure to correct
a deficiency
with the emergency
diesel
generator starting air system,
paragraph
3.
An
NRC identified non-cited violation was identified regarding the failure to
properly implement plant procedures
for equipment control, paragraph 2.a.(l)
The content of operator logs describing
a failure in the charging/safety
injection system
was considered
to be deficient,
paragraph
2,a.
921014015l
92100l
ADOCK 05000400
8
The licensee's
administrative controls for ensuring reliable decay heat
removal during outages
were considered
to be good
and
had implemented industry
recommendations
to reduce the potential for core
damage
events
during outage
activities,
paragraph
5.
Improvement
was noted in technical
support center
command
and control
and in
the involvement of the accident
assessment
team during the annual
emergency
drill, paragraph
6.
REPORT DETAILS
1.
Persons
Contacted
Licensee
Employees
- J. Collins,- Manager,
Operations
J. Cribb, Manager,
gual'ity Control
- C. Gibson,
Manager,
Programs
and Procedures
- C. Hinnant,
General
Manager,
Harris Plant
D. Knepper,
Project Engineer,
Nuclear Engineering
Dept.
B. Meyer,
Manager,
Environmental
and Radiation Monitoring
- T. Morton, Manager,
Maintenance
- J. Hoyer,
Manager,
Project Assessment
- J. Nevill, .Manager,
Technical
Support
- C.'Olexik, Manager,
Regulatory
Compliance
A. Powell,
Manager,
Harris Training Unit
H. Smith,
Manager,
Radwaste
Operation
- G. Vaughn,
Vice President,
Harris Nuclear Project
- E. Willett, Manager,
Outages
and Modifications
W. Wilson, Manager,
Spent Nuclear
Fuel
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.
2.
Review of Pl'ant Operations
(71707)
~ The plant began this inspection period in power operation
(Mode 1).
On
September
12,
1992,
a plant shutdown
was
commenced for a scheduled
refueling outage.
At II:00 p.m.
on September
13, the plant was cooled
down to the cold shutdown
(Mode 5) condition.
The plant remained
in
cold shutdown for the duration of this inspection period.
a.
Shift Logs
and Facility Records
The inspector
r'eviewed records -and discussed
various entries with
operations
personnel
to verify compliance with the Technical
Specifications
(TS)
and the licensee's
administrative procedures.
The following records
were reviewed:
Shift Supervisor's
Log;
Outage Shift Manager'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 generally readable,
well
organized,
and provided sufficient information on plant status
and
events.
However,
a review of the control
room logs associated
with the miniflow drain valve failure discussed
in LER 91-08 were
found to be deficient.
The shift foreman'.s
and reactor operator's
logs were reviewed for the broken drain line event of March 22,
1991.
The shift foreman's
log did not document the drain line
failure or the dumping of hundreds of gallons of 'water
on the
floor.
The reactor operator's
log only documented
closing two
valves
because
of a leak.
The failure to adequately
document the
event in the operating logs in this specific case
was considered
to be
a weakness.
During a routine tour of the control
room on August 28,
1992, the inspector
observed
licensee activities regarding
the failure of several
ESFAS valves to reposition during
surveillance testing.
Licensee
personnel
determined that
an
equipment clearance
had
been established
on August. 14,
1992,
(clearance
OP-92-0988)
which removed
a fuse in the solid
state protection logic circuitry to deenergize
an inoperable
blowdown isolation valve (1BD-7).
The
removed fuse also unintentionally affected
11 other steam
generator
sample
and
blowdown valves.
These
types of
isolation valves
are administratively controlled by the
licensee
as safeguards
systems
isolation valves in closed
systems
and
have associated
action statements
similar to
containment isolation valves.
Licensee
personnel
replaced
the fuse
and disabled
valve
instead.
Procedure
OMM-014, Operations-Operation
of the
Clearance
Center,
contains
guidance to ensure that equipment
is correctly and safely removed
from service.
Section
5. 1.2
of this procedure directs the clearance
preparer to review
control wiring diagrams
(CWDs)
as appropriate to establish
the required
equipment lineup to isolate the affected
component.
The inspector reviewed the applicable
CWD
(CAR
2166
B-401 sheet
1191) to locate the removed fuse.
The
inspector considered
that the
CWD contained
clear references
to other
CWD's and components
affected
by the removal of the
fuse.
Therefore,
the inspector
concluded that appropriate
CWD's were not properly reviewed prior to initiating the
equipment clearance
as required
by plant procedures.
Although several
valves were affected
by the equipment
clearance,
the
SSPS contained internal
redundant circuitry
which would have actuated
the components if required.
The
SSPS feature to isolate the steam generator- blowdown and
sample valves
was not required
by the TS.
This information
lessened
the safety significance of the issue.
-In response
to this event,
the licensee
conducted real-time training for
all shift operators
to stress
the importance of thorough
drawing reviews for the establishment
of equipment
clearances.
This
NRC identified violation is not being
cited because criteria specified in Section
V.A of the
NRC
Enforcement Policy were satisfied.
NCV (400/92-17-01):
Failure to properly implement plant
procedures
for equipment control,.
b.
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;
reactor containment building; fuel handling, building; emergency
service water building; battery rooms; electrical
switchgear
rooms;
and the technical
support center.
During these tours,
the following observations
were made:
(1)
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.
(2)
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 turnovers
on various occasions
to verify the
continuity of plant status,
operational
problems,
and other
pertinent plant information during these turnovers.
(3)
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.
i
(5)
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 vehi'cles;
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.
(6)
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.
The inspectors
found plaAt housekeeping
and. material condition of
safety related
components
to be good.
The licensee's
adherence
to
r'adiological controls, security controls, fire protection
requirements,
and
TS requirements
in these
areas
was satisfactory.
c.
Review of Nonconformance
Reports
Adverse Condition Reports
(ACRs) were reviewed to verify the
following: .TS were complied with, corrective actions
and generic
items were identified and items were reported
as required
by
10 CFR 50 '3.
Surveil'lance 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:
=
~ HST- I0027
~'ST- I0044
B Narrow Range
Level
(L-0484) Calibration
Calibration of Nuclear Instrumentation
System
Power
Range
~ HST-10169 Nuclear Instrumentation
System
Source
Range
N31 Operational
Test
~ HST- I0268 Lo-Lo TAVG P- 12 Interlock (T-0412) Operational
Test
~ OST-1007
CVCS/SI System Operability quarterly Interval
~ OST-1107
ECCS Flow Path
and Piping Filled Verification Honthly
Interval
~ OST-1801
ECCS Throttle Valve,
CSIP
and Check Valve Verification 18
Nonth Interval
~ OST-1823
1A-SA Emergency Diesel
Generator
18 Honth Operability. Test
~ EPT-189
Alternate Mini Flow Relief Valves (ICS-744
and ICS-755) Full
Flow Test
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.
While observing the performance of procedure
OST-1823,
on September
14,
1992, it was not'ed that the diesel
had failed to meet its start time
requirement.
A work request
was initiated to correct this deficiency.
Subsequently,
the diesel
was restarted for trip testing during which it
failed to trip within the required'time
band during
a loss of potential
transformer trip test.
Another work request
was initiated to correct
this deficiency.
During troubleshooting,
the licensee
found that
a shuttle valve, which
pressurizes
the governor oil booster for opening the fuel racks
on the
EDG,
had failed.
The failure of the fuel racks to open subsequently
resulted
in the slow diesel
generator start.
During inspection of the
shuttle valve, the valve control air ports were found to be clogged
which indicated debris in the control air system.
During
troubleshooting for the failed trip test,
the orifice for bleeding off
the control air was found to be partially obstructed., This also
indicated debris in the control air system.
The control air and
starting air flasks were subsequently
blown down to ensure
proper air
quality.
Discussion with licensee
personnel
disclosed that the diesel
generator
air system
had previously experienced
a high moisture problem during the
last refueling outage.
After refilling the air flasks, the
dew point
was*found to be 80 degrees
F.
and the control air filters were found
partially filled with water and Neolube residue.
The high dew point and
water in the control air filters indicated that the air drying/moisture
removal
systems
were inadequate.
Further review found that
a plant
modification,
PCR-3995,
EDG Starting Air Dryer Drains,
had
been
initiated on November
14,
1988,
becau'se
the air drying system
was
becoming saturated
and
was not able to perform=properly.
This
modification, which has not been
implemented to date, will add
a water
cooled compressor
aftercooler
and a,moisture
separator
in order to
reduce the moisture content of the compressed
air.
This in turn will
allow the air dryer desiccant
towers .to adequately
dry the air going to
the air flask.
The inspectors
reviewed Generic Letter 88-14,
Instrument Air Supply
System
Problems Affecting Safety-Related
Equipment,
and the licensee's
subsequent
response.
The generic letter requested
that the licensee
verify that the air quality was consistent
with the manufacturer's
recommendations.
In the licensee's
response
dated
February 3,
1989, the
licensee
stated that for the diesel
generator starting air system,
the
actual
measured
dew point in the tanks
was
above desirable levels
due to
moisture collecting in low points in the air drying towers.
The
licensee
stated that plant modifications would be pursued to correct
this problem.
10 CFR 50, Appendix B, Criteria XVI, requires that measures
shall
be
established
to assure that conditions
adverse
to quality are promptly
identified and corrected.
Although Generic Letter 88-14 notified the
industry of potential air system
problems
and the l.icensee identified
problems with maintaining
dew point levels after receiving the generic
letter, the system deficiencies
were not corrected
promptly.
As
a
result,
poor air quality contributed to the diesel
generator start
failure on, September
14,
1992.
The failure to promptly correct this
deficiency is considered
to be
a violation.
Violation (400/92-17-02):
Failure to correct
a deficiency with the
starting air drying system.
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,
as required,
were issued
and being followed;
quality control personnel
were available for inspection activities
as
required;
and
TS requirements
were being followed.
Maintenance
was observed
and work packages
were reviewed for the
following maintenance activities:
~ Diesel generator
shuttle valve replacement
due to air blockage.
~ Troubleshooting of the diesel
generator
loss of potential transformer
emergency trip due to orifice blockage.
~ Troubleshooting
the failure to get
a first out annunciator
on
a manual reactor trip signal.
~ Troubleshooting
the containment
spray
pump supply breaker closing
spring motor failure.
The performance of work was satisfactory with proper documentation
of
removed
components
and independent verification of the reinstallation.
No violations or deviations
were identified.
Reliable
Decay Heat
Removal
During Outages
(Closed)
TI 2515/113:
This special
inspection
was performed to review
licensee activities during the refueling outage
which have the potential
to cause
a loss of decay heat
removal capabilities.
The inspectors
interviewed outage personnel,
and reviewed the outage
schedule,
licensee
procedures
and administrative controls,
and the licensee's
shutdown risk
assessment
report of the outage
schedule.
In addition to TS
requirements,
administrative controls in the following plant procedures
were reviewed:
'
PG0-054,
Control of Plant Activities During Reduced
Inventory
Conditions
~ PG0-060,
Outage Risk Management
Policy and Principles
~ PLP-700,
Outage
Management
~ EM-005, Temporary
Power for Bus Outages
~
NUMARC 91-06, Guidelines for Industry Actions to Assess
Shutdown
Management
~ NUREG-1449 (Draft), Shutdown
and
Low-Power Operation of Commercial
Nuclear
Power Plants in the United States
The inspectors
found that administrative controls properly addressed
and
identified any operations
which potentially jeopardized
decay heat
removal capability during the outage.
The licensee
did not plan to
establish
a reduced
RCS inventory condition until fuel
was completely
off-loaded from the core.
Other special tests
or operations
which might
also affect
DHR capability were not found.
Appropriate actions
had
been
included in the outage
schedule to maintain at least
one onsite
power
source,
with corresponding
emergency diesel
generator
and
distribution system,
and
one offsite power source available at all
times.
The licensee
planned to use non-standard
electrical
lineups to
backfeed
power through the main transformers
and unit auxiliary
transformers
when maintenance
was performed
on the startup transformers.
Also, the licensee
planned to cross-tie
the general
service
bus
(1-4A)
sections
to allow maintenance
on
an electrical
bus.
These electrical
lineups were addressed
in the
FSAR and designed
to carry the electrical
load.
Operating
procedures
specified the actions required to accomplish
the electrical
lineups.
Individual loads which would receive electrical
power from a temporary
source
were addressed
by approved
procedures
or
issuance
of temporary'odifications
which undergo safety reviews.
The
inspector also verified that abnormal
and emergency
procedures
were
available which addressed
loss of electrical
power and improper
automatic action of the emergency
load sequencer.
The licensee's
emergency diesel
generator
receives
a field flashing source
from safety
related
125
VDC emergency distribution panels.
The inspector verified
that licensee
personnel
would declare
the diesel
inoperable if these
sources
were
removed
from service.
The licensee's
shutdown risk management
program was modeled after the
NUMARC 91-06 guidance
and included provisions for identification of
higher risk evolutions
and key safety functions/equipment,
defense-in-
depth,
contingency planning, training,
and outage risk assessment
l
h
reviews.
The inspector
was informed that the licensee's
program
contained
in procedure
PLP-700
had not been officially approved at the
start of the current outage.
Nevertheless
licensee
management
committed
to comply with the administrative controls provided in the program until
the program was approved.
The risk assessment
review was conducted
by
a
multi-disciplined group.
The resulting report included recommendations.
for additional training to appropriate plant personnel
regarding
defense-in-depth
and risk assessment
details.
This group identified the
planned
process of switching electrical
power for the "B" spent fuel
cooling
pump to be
a higher risk evolution due to plant conditions
(one
spent fuel cooling pump available)
and previous
problems
experienced
by
plant personnel
during this process.
Also, implementation of
containment integrity was conservatively
recommended
to be available
during lifting of the reactor vessel
head.
Changes
to .the schedule
were
implemented to include these
recommendations.
Changes
to the outage
schedule
which involve a schedule logic change
must undergo
a review by a team consisting of an
STA, shift outage
manager,
maintenance shift manager,
and technical
support shift manager.
The
PNSC must approve
any significant change
which results
in a higher
risk evaluation or
a reduction in the defense-in-depth.
The outage
schedule
provided for redundancy for key safety functions
by
.
utilization of the spare
CSIP,
CCW pumps
and
by availability of NSW
during
ESW maintenance.
The'scheduling
checklist utilized for schedule
creation
included detailed
requirements
for electrical power,'ore
cooling, support
system availability, spent fuel cooling,
makeup
capability,
RCS pressure
control,
and controls for reduced
RCS inventory
conditions.
The licensee's
actions regarding
RCS reduced
inventory operation
and
potential
loss of decay heat
removal capability were previously reviewed
in
NRC Inspection
Report 50-400/91-09.
The inspectors
found that the licensee's
administrative controls for
ensuring reliable decay heat
removal during outages
were good
and
implemented
industry recommendations
to reduce the potential for core
damage
events during outage activities.
Annual
Emergency Drill (71707)
On August 27,
1992, the annual
emergency drill was conducted
by the
licensee to verify the effectiveness
of the Radiological
Emergency
Response
Plan
and implementing procedures.
Details of the drill,
including the results of critiques held,
are discussed
in NRC Inspection
Report 50-400/92-16'.
Improvement
was noted in TSC
command
and control which strengthened
communication
and limited the noise level.
Also, the involvement of the
'accident
assessment
team
was considered
to be good.
Review of Licensee
Event Reports
(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
(Closed)
LER 91-08:
This
LER reported that the high head safety
injection system
was inoperable
due to
a failure of the system's
alternate miniflow lines.
This event
was previously discussed
in
NRC Inspection
Report 50-400/92-15.
Additional followup of this
event
was performed
as documented
in
NRC Inspection
Report 50-
400/92-201.
Due to continuing
NRC concerns that water
hammer
events
had caused relief valve damage
and drain line failure,
licensee
management
committed to test the system during this
refueling outage.
The inspector
observed this testing
on September
17,
1992.
Although severe
water
hammer
was not observed
during the train
A
or train
B miniflow testing, relief valve chattering
was observed
with the "B" train being more severe.
The "A" charging
pump
and
miniflow system test
was satisfactory with minor relief valve
chatter for about four seconds
on initial isolation valve opening
and closing.
The "8" charging
pump
and miniflow system exhibi,ted
increased relief valve chattering
on initial isolation valve
opening.
The chattering continued with the isolation valve fully
open.
This resulted
in the rupturing of the relief valve bellows
and water leakage
onto the floor. It appeared
to the inspector
that the relief valve would have continued to chatter
as long as
the
pump discharge
pressure
was
above the relief valve setpoint.
The licensee
had instrumented
the piping system with vibration
monitoring equipment
and pressure
transmitters,
and videotaped
the
actuations.
A detailed failure analysis
was not available at the
end of this inspection period.
The Crosby relief valve technical
manual
stated that troubles
encountered
with relief valves often vitally affect the life,
operation
and performance of the valve and should
be corrected
as
soon
as possible.
The manual listed valve chattering
as
one of
these troubles
and stated that chattering
could be caused
by
a
restricted'inlet to the valve.
The present
design which uses
an
isolation valve is contrary to the manufacturer's
recommendations
in that the isolation valve restricts
the inlet to the relief
valve while the isolation valve is opening
and closing.
It
appeared
to the inspector that the chattering
was caused
by the
choking of the relief valve inlet by the isolation valve.
The
inspector believed that valve chattering contributed to the
various valve failures in the past
and to the "B" train drain line
piping failure and the drain line pipe cracking observed
on the
"A" train.
10
A review of previous
system testing
was performed.
During this
review it was disclosed that the miniflow system
had never
been
're-operationally
tested
by the licensee
to insure the adequacy of
the original design provided
by Westinghouse.
It was also noted
that while performing the safety injection miniflow system testing
in the past,
the relief valves
had
been
removed
from the system
and orifice plates
were installed instead.
The relief valves were
then
bench tested for setpoint
accuracy.
Although system
components
had
been tested,
the system
had never
been fully
functionally tested.
Following the relief valve failures
and the broken drain line of
Harch 22,
1991, the licensee initiated
LER 91-08.
This
LER
documented
the corrective actions
taken to prevent recurrence of
the event.
The
LER identified water
hammer
caused
by inadequate
venting of the system
as the root cause of the system
damage.
Although the inspectors
questioned
the use of relief valves in
this system,
the licensee
did not perform further testing to
insure
system operabilitp.
Had
a functional test of this system,
similar to that performed
on September
17,
been
performed
following the event in 1991, the valve chattering
problem would
have
been identified much earlier.
The licensee's
corrective
action for this event failed to identify that
a system functional
test
had never
been
performed to verify the adequacy of the
design.
10 CFR 50 Appendix B, Criterion XVI requires that conditions
adverse
to quality are promptly identified and corrected.
Contrary to this requirement,
the licensee failed to properly test
the miniflow system to ensure that corrective actions
were
adequate
and to ensure
proper system operation.
This failure is
considered
to be
an apparent violation.
Apparent Violation (400/92-17-03):
Failure to promptly identify
and correct
an adverse, condition.
As discussed
in the
LER, licensee
personnel
were
aware that the
miniflow system drain line had fallen off on March 22,
1991,
and
had generated
ACR 91-108.
However, this
ACR was categorized
as
non-significant
when reviewed.
Since only significant
ACRs
receive root cause
investigations,
the failure to identify ACR 91-
108 as
a significant condition delayed the investigation.
The
ACR
was not identified as significant until the miniflow system relief
valve testing failures occurred
on April 3,
1991,
and
by then the
drain line had already
been
rewelded.
Therefore,
a root cause
analysis of the weld failure could not be performed.
Procedure
PLP-002, Corrective Action Program,
Section 5.2.5, requires
the
CAP manager to determine if the condition is significant per
attachment
3.
Attachment 3, item 1, lists
a condition that
has or
reasonably
could cause
a loss of safety
system function
as
a
significant condition.
Since the drain line on
a safety
system
had broken off, which would have caused
a significant loss of flow
b.
c ~
from the
"B," charging/safety
injection pump, this
ACR should
have
been categorized
as significant.
The failure to follow the
corrective action program procedure is considered
to be another
example of the apparent violation discussed
above.
During a design review of the charging/SI
system,
the inspectors
noted that due to the high alternate miniflow recirculation,
flow may not be available over the full range of RCS design
pressures
(2575
PSIG)
and
may not meet the minimum flow
requirements
depicted
in the
FSAR Small
Break Safety Injection
Flowrate Analysis, Figure 15.6.5-56.
The licensee
reviewed this
concern
and stated that adequate
flow would still be available to
satisfy the safety analysis.
When this concern
was raised,
the
inspectors
were informed that the licensee
had already initiated
a
PCR to remove the miniflow relief valves during the
1992 refueling
outage.
The modification package
was onsite for preliminary
review before
the testing
on September
17,
1992.
Prior to this
testing,
the licensee
had determined that
a modification of the
design
was desired,
and the modification was
scheduled
to be completed prior to startup
from the present
refueling outage.
(Open)
LER 92-12:
This
LER reported
a failure of the "B"
emergency
bus undervoltage
relay which caused
an entry into TS 3.0.3.
The licensee
replaced
the faulty relay
and revised
appropriate testing
procedures
to ensure
the relays trip as
required.
Plant modifications are being planned to ensure that
target flags properly actuate with the relays
and instructions
regarding relay removal
from service will be clarified. The
LER
will remain
open pending completion of this additional action.
(Open)
LER 92-13:
This
LER reported that
a
TS surveillance
requirement to sample the fuel oil day tank was not performed
following emergency diesel
operation.
The licensee
plans to train
applicable
personnel
and revise administrative
procedures
to
prevent recurrence of this event.
Individuals involved in the
event will be counseled.
Exit Interview (30703)
The inspectors
met with licensee
representatives
(denoted
in paragraph
1) at the conclusion of the inspection
on September
25,
1992.
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.
12
Item Number
400/92-17-01
400/92-17-02
400/92-17.-03
Descri tion and Reference
NCV:
Failure to properly implement
plant procedures
for equipment
control, paragraph 2,a.(l).
VIO:
Failure to correct
a
deficiency with the
EDG starting air
dryer system,
paragraph
3.
Apparent Violation:
Failure to
promptly identify and correct
an
adverse
condition, paragraph
7.a.
and Initi al i sms
ACR
CFR
CSIP'VCS
CWD
ESFAS-
LER
NRC
NUHARC-
NUREG-
PNSC
RCS/RC-
SSPS
TAVG
TS
VDC
Adverse Condition Report
Corrective Action Program
Component Cooling Water
Code of Federal
Regulations
Charging Safety Injection
Pump
Chemical
Volume Control
System
Control Wiring Diagrams
Direct Current
Decay Heat
Removal
Emergency
Core Cooling System
Emergency Diesel
Generator
Engineered
Safety Features
Actuation System
Emergency Service Water
Final Safety Analysis Report
Licensee
Event Report
Non-Cited Violation
Nuclear Regulatory
Commission
Normal Service
Water
Nuclear Nanagement
and Resources
Council
NRC Technical
Report Designation
Plant
Change
Report
Plant Nuclear Safety Committee
Pounds
per Square
Inch Gage
Safety Injection
Solid State Protection
System
Average
RCS Temperature
Technical Specification
-Technical
Support Center
Volts Direct Current
Violation